Determination of LiF:Mg,Ti and LiF:Mg,Cu,P TL efficiency for X-rays and their application to Monte Carlo simulations of dosemeter response

The simulation of response of a new passive area dosemeter for measuring ambient dose equivalent H*(10) for photons has been performed using the Monte Carlo code MCNP and experimentally determined responses of LiF:Mg,Ti and LiF:Mg,Cu,P thermoluminescent (TL) detectors for hard-filtered X-ray spectra from 20 to 300 keV and for 137Cs and 60Co gamma radiation. Relative TL efficiency for both types of detectors, determined in experiments with bare detectors and similar Monte Carlo simulations, compared favourably with prediction of microdosimetric models for proposed microdosimetric target sizes in the range of 20-40 nm. The concluding verification experiment showed small deviations between measured and simulated dosemeter energy response values in the range of a few percent.     

Calculated angular responses of an RPL dosemeter to photon and beta radiation

The calculated dose equivalent response as a function of the angle has been examined for the radiophotoluminescent (RPL) glass dosemeter that was exposed to narrow series X-ray, N-60, N-80, N-100, N-150, N-200, N-250, N-300, photon sources ((60)Co and (137)Cs) and beta-ray emitter ((90)Sr/(90)Y) while mounted on an ISO water slab phantom. The angular dose equivalent responses H(p)(10) and H(p)(0.07) were calculated using the Monte Carlo MCNPX code. The RPL dosemeter and the phantom were rotated in the horizontal and vertical planes from a variety of angles of interest. The results were compared with the experimental data. Good agreement was found between the measured and calculated values of the relative dose equivalent angular responses of the RPL dosemeter.     

Microdosimetric interpretation of the photon energy response of LiF:Mg,Ti detectors

Photon energy response of MTS-N (LiF:Mg,Ti) detectors (TLD Poland) and of MTS-N detectors sensitised with 200 Gy of 60Co gamma rays, followed by UV irradiation (sMTS-N), has been determined using X rays with narrow energy spectra, in the energy range from 20 to 300 keV. The over-response of LiF:Mg,Ti detectors for X rays (relative TL efficiency eta = 1.1) can be explained as an ionisation density effect. Low energy X rays produce short electron tracks, which locally deposit a high radiation dose and, consequently, lead to an enhanced (supralinear) response. This over-response has not been observed in sensitised MTS-N where supralinearity in the response after gamma ray doses above 1 Gy is not seen. Using the dose-response curves measured for MTS-N detectors after 137Cs gamma ray irradiation and local doses calculated using Monte Carlo generated electron tracks, it was possible to predict the relative TL effectiveness for different X ray energies. The calculation procedure can be applied to predict the photon energy response of LiF:Mg,Ti detectors in an arbitrary photon field.     

The influence of the type of filling gas on the response of ionisation chambers to a mixed high-energy radiation field

Radiation protection dosimetry in radiation fields behind the shielding of high-energy accelerators such as CERN is a challenging task and the quantitative understanding of the detector response used for dosimetry is essential. Measurements with ionisation chambers are a standard method to determine absorbed dose (in the detector material). For applications in mixed radiation fields, ionisation chambers are often also calibrated in terms of ambient dose equivalent at conventional reference radiation fields. The response of a given ionisation chamber to the various particle types of a complex high-energy radiation field in terms of ambient dose equivalent depends of course on the materials used for the construction and the chamber gas used. This paper will present results of computational studies simulating the exposure of high-pressure ionisation chambers filled with different types of gases to the radiation field at CERN's CERN-EU high-energy reference field facility. At this facility complex high-energy radiation fields, similar to those produced by cosmic rays at flight altitudes, are produced. The particle fluence and spectra calculated with FLUKA Monte Carlo simulations have been benchmarked in several measurements. The results can be used to optimise the response of ionisation chambers for the measurement of ambient dose equivalent in high-energy mixed radiation fields.     

Calibration of a 7.6 cm x 7.6 cm (3 inch x 3 inch) sodium iodide gamma ray spectrometer for air kerma rate

An experimental procedure is described for converting a gamma ray spectral measurement from a 7.6 cm x 7.6 cm (3 inch x 3 inch) sodium iodide (NaI) detector to air kerma rate. The calibration procedure involves measuring the energy deposited in the detector using 10 radioactive sources of known activity covering an energy range from 60 keV to 1,836 keV. For each of the 10 sources, gamma ray spectra were measured with the source at different angles to the detector axis. The total energy deposited in the detector for the ten sources was confirmed by Monte Carlo calculations. The spectra measured at different angles were combined to produce a spectrum that would represent a homogeneous semi-infinite source of radiation. The resultant spectrum was then subdivided into 10 energy regions. Based on the known air kerma rates due to the sources, a calibration coefficient was calculated for each of the 10 energy regions. These calibration coefficients could then be used to convert the energy deposited in the 10 regions of an unknown spectrum to air kerma rate. The calibration procedure was confirmed by comparing the results from the detector with those from calibrated collimated beams of 137Cs and 60Co. A comparison of measurements using a calibrated pressurised ionisation chamber with those from a similar Nal spectrometer in Finland provided additional confirmation of the calibration procedure.     

The response of lif thermoluminescence dosemeters to photon beams in the energy range from 30 kV x rays to 60Co gamma rays

The energy response of standard (TLD-100) and high-sensitivity (TLD-100H) LiF thermoluminescence dosemeters (TLDs) has been studied for photon beams with mean energies from about 25 keV to 1100 keV. Canadian primary standards for air kerma were used to establish the air kerma rates for each of the photon beams. TLDs were mounted in a PMMA holder and the air kerma response was measured as a function of energy. The EGSnrc Monte Carlo code was used to model the TLD holder and calculate the absorbed dose to the TLD chip per unit air kerma for each beam. The measured and calculated results were combined to obtain the intrinsic dose response of the TLD chip. Broadly, our results are consistent with existing data, which show a marked difference in the energy dependence of the two materials. However, the precision of our measurements (standard uncertainty of about 0.6%) has permitted the identification of features that have not been noted before. In particular, the energy dependence of the two materials is quite different in the important energy region delimited by 137Cs and 60Co gamma rays.     

Characteristics of a 85Kr beta-particle source applied in Series 1 reference irradiations of DIS-1 direct ion storage dosemeters

Characteristics necessary to specify an ISO 6980 Series 1 reference radiation field were determined for a commercially available 85Kr beta-particle source, using a BEAM EGS4 Monte Carlo code. The characteristics include residual maximum beta energy, E(res), and the uniformity of the dose rate over the calibration area. The E(res) and the uniformity were also determined experimentally, using an extrapolation ionization chamber (EC) and a 0.2 cm3 parallel plate ionization chamber, respectively. The depth-dose curve measured with the EC gave a value 0.62 MeV for the E(res). Series 2 90Sr + 90Y and Series 1(85) Kr beta-particle sources calibrated for H(p)(0.07) at the secondary standard dosimetry laboratory (SSDL) of STUK were used to determine the energy and angular responses of DIS-1 direct ion storage dosemeters. The averaged zero angle H(p)(0.07) responses to the 90Sr + 90Y and 85Kr reference radiations were 135 and 80%, respectively. The responses were normalized to 100%, H(p)(0.07) response to 137Cs photon radiation.     

Non-destructive method of characterisation of radioactive waste containers using gamma spectroscopy and Monte Carlo techniques

During the decommissioning of the SATURNE accelerator at CEA Saclay (France), a number of concrete containers with radioactive materials of low or very low activity had to be characterised before their final storage. In this paper, a non-destructive approach combining gamma ray spectroscopy and Monte Carlo simulations is used in order to characterise massive concrete blocks containing some radioactive waste. The limits and uncertainties of the proposed method are quantified for the source term activity estimates using 137Cs as a tracer element. A series of activity measurements with a few representative waste containers were performed before and after destruction. It has been found that neither was the distribution of radioactive materials homogeneous nor was its density unique, and this became the major source of systematic errors in this study. Nevertheless, we conclude that by combining gamma ray spectroscopy and full scale Monte Carlo simulations one can estimate the source term activity for some tracer elements such as 134Cs, 137Cs, 60Co, etc. The uncertainty of this estimation should not be bigger than a factor of 2-3.     

Evaluation of the dose rate distribution for an air-type 60Co irradiation facility

The dose rate distributions in the 29,000-Ci 60Co irradiation facility in National Tsing Hua University were investigated by measurements and calculations. The dose rate measurements were performed using radiochromic dye films and an Exradin A2 ion chamber mounted on a PC-controlled motorised vertical translation stage. The calculations were made by using the three-dimensional point kernel code QAD-CGGP with detailed source composition and geometry modelling. The scattered gamma rays from the walls of the irradiation cell were also evaluated by using the Monte Carlo code MCNP.     

Design and calibration of a gamma ray transmission density probe

A dual probe gamma ray transmission gauge has been designed in LNETI for the measurement of soil porosities and bulk densities. The gauge makes use of a 4 cm GM counter and of a ¹³⁷Cs source, and can be operated at a source-detector distance of either 20 or 30 cm.The gauge performance and main parameters were studied experimentally as well as by means of a computer simulation. The experiments were carried out both in the laboratory and on the field. Monte Carlo simulations were performed for both Cs and Co sources and the two source-detector distances. Reported here are calibration curves for sensitivity, relative statistical errors and resolution. Good agreement has been found between simulated and experimental data.     

Shielding evaluation of a medical linear accelerator vault in preparation for installing a high-dose rate 252Cf remote afterloader

In support of the effort to begin high-dose rate 252Cf brachytherapy treatments at Tufts-New England Medical Center, the shielding capabilities of a clinical accelerator vault against the neutron and photon emissions from a 1.124 mg 252Cf source were examined. Outside the clinical accelerator vault, the fast neutron dose equivalent rate was below the lower limit of detection of a CR-39 etched track detector and below 0.14 +/- 0.02 muSv h(-1) with a proportional counter, which is consistent, within the uncertainties, with natural background. The photon dose equivalent rate was also measured to be below background levels (0.1 muSv h(-1)) using an ionisation chamber and an optically stimulated luminescence dosemeter. A Monte Carlo simulation of neutron transport through the accelerator vault was performed to validate measured values and determine the thermal-energy to low-energy neutron component. Monte Carlo results showed that the dose equivalent rate from fast neutrons was reduced by a factor of 100,000 after attenuation through the vault wall, and the thermal-energy neutron dose equivalent rate would be an additional factor of 1000 below that of the fast neutrons. Based on these findings, the shielding installed in this facility is sufficient for the use of at least 5.0 mg of 252Cf.     

Application of voxel phantoms in whole-body counting for the validation of calibration phantoms and the assessment of uncertainties

This article is dedicated to the application of voxel phantoms in whole-body counting calibration. The first study was performed to validate this approach using IGOR, a physical phantom dedicated to fission and activation product (FAP) measurement, and a graphical user interface, developed at the IRSN internal dose assessment laboratory, called OEDIPE (French acronym for the tool for personalised internal dose assessment) associated with the Monte Carlo code MCNP. The method was validated by comparing the results of real measurements and simulations using voxel phantoms obtained from CT scan images of IGOR. To take this application further, two studies were carried out and are presented in this article. First, a comparison was made between the IGOR voxel based phantom (IGOVOX) and a voxel human body (Zubal Phantom) to confirm whether IGOR could be considered as a realistic representation of a human. Second, the errors made when considering sources homogeneously distributed in the body were assessed against real contamination by taking into account the biokinetic behaviour of the radioactive material for two modes of exposure: the ingestion of (137)Cs in soluble form and the inhalation of insoluble (60)Co several days after acute incorporation.     

Evaluation of the photoneutron field produced in a medical linear accelerator

The doses and spectra of photoneutrons produced in a medical linear accelerator with photon energies of 10 and 15 MV were evaluated. The Monte Carlo code, MCNPX, was used to simulate the transport of these photoneutrons around the head for 10 and 15 MV photons. The fully-described geometry of the accelerator head was used in this calculation. The photoneutron energy spectra and doses for various photon field sizes were calculated at each of 20 positions. The results indicate that the maximum dose equivalents are observed in 20 x 20 cm(2) case among photon fields. It was found the neutron average energy at isocenter for a 0 x 0 cm(2) field is 0.38 MeV for 10 MV and is 0.45 MeV for 15 MV. The neutron doses at 10 positions around the head in the treatment room of the operation facility at 10 and 15 MV were measured using the bubble detectors. Measurements were compared with the calculations under the same geometry in the experiment. It was found that the majority of the calculated results agreed to within the standard deviations of the measurements. These above results can be applied in the verification of maximum allowed neutron leakage percentage of treatment dose defined in the IEC. We have been employing them to derive the empirical formula for neutron dose equivalent level at the maze entrance of medical accelerator treatment rooms in a study that is still underway.     

An experimental test of computed responses of bismuth germanate to gamma rays

Experimental spectra of the γ-rays from ⁶⁰Ca, ²²Na, ²⁰⁷Bi and ¹³⁷Cs were used to calibrate and obtain spectra from a 3.17 cm diameter by 0.65 cm thick BGO scintillation spectrometer. These spectra were compared to the results of detailed Monte Carlo simulations. Tables of coefficients for polynomials to give absorption coefficients are presented.     

Monte Carlo study of a 60Co calibration field of the Dosimetry Laboratory Seibersdorf

The gamma radiation fields of the reference irradiation facility of the Dosimetry Laboratory Seibersdorf with collimated beam geometry are used for calibrating radiation protection dosemeters. A close-to-reality simulation model of the facility including the complex geometry of a (60)Co source was set up using the Monte Carlo code MCNP. The goal of this study is to characterise the radionuclide gamma calibration field and resulting air-kerma distributions inside the measurement hall with a total of 20 m in length. For the whole range of source-detector-distances (SDD) along the central beam axis, simulated and measured relative air-kerma values are within +/-0.6%. Influences on the accuracy of the simulation results are investigated, including e.g., source mass density effects or detector volume dependencies. A constant scatter contribution from the lead ring-collimator of approximately 1% and an increasing scatter contribution from the concrete floor for distances above 7 m are identified, resulting in a total air-kerma scatter contribution below 5%, which is in accordance to the ISO 4037-1 recommendations.     

Comparison of two extremity dosemeters based on LiF:Mg,Cu,P thin detectors for mixed beta-gamma fields

Two types of thin LiF:Mg,Cu,P detectors, GR-200F and MCP-Ns, have been characterised for use in the design of an extremity dosemeter for mixed beta-photon radiation fields. Both detectors consist of an extremely thin layer of sensitive material with effective thicknesses of 5 and 8 mg cm(-2), respectively, held in a 5 mg cm(-2) PVC ring holder. Dosimetric performance was analysed according to the ISO 12794 standard and compared with 240 mg cm(-2) TLD-100 measurements. In particular, the energy response was obtained for ISO narrow X-ray spectra, (137)Cs, (60)Co, (204)Tl and (90)Sr/(90)Y. From these measurements a mean calibration factor was calculated to estimate H(p)(0.07). Subsequently, the performance of the dosemeters was checked for a set of 10 different mixed photon and beta-photon fields. The study shows that the proposed dosemeters can estimate H(p)(0.07) in a wide range of mixed beta-photon fields with a maximum deviation from the given dose of 30% and an overall uncertainty of the order of 25% (k = 1). However, the results also highlight a large variability among the different thin detectors and, thus, the standard TLD-100 material is recommended whenever the workplace does not include low-energy beta radiation.     

An investigation of energy spectrum and lineal energy variations in mega-voltage photon beams used for radiotherapy

Megavoltage photon beams are routinely used for external-beam radiotherapy. Recently, new treatment modalities based on dynamic and intensity modulated (IM), beam delivery systems are increasingly used in clinical practice. The purpose of this work is to investigate the energy spectrum and microdosimetric features of these photon beams. A Monte Carlo technique was first used to simulate beam lines of medical accelerators and to compute photon fluence and spectrum per unit dose-to-water inside the irradiated medium. Subsequently, a track structure code was used to compute the lineal energy and its distribution in a 1 micron sphere based on the individual photon spectrum. Results showed that the low energy photon component varied significantly with field size and location within the field due to the presence of the scattered photons. The calculated dose-mean lineal energy ranged from 2.3 keV.micron-1 at a depth of 1.5 cm along the central axis of a 4 cm x 4 cm field, to 3.7 keV.micron-1 at a depth of 20 cm of the field edge in a 10 cm x 10 cm field. In the tested IM fields at a depth of 10 cm, ranged from 2.6 to 3.5 keV.micron-1, which was inversely related to the dose intensity in the field. The values for the clinical photon beams were also significantly greater than that of a reference 60Co beam (1.8 keV.micron-1 from the calculation). The beam quality factor was estimated to vary within 20% due to the change of the energy spectrum for the radiotherapy photon fields.     

Response of alanine and radio-photo-luminescence dosemeters to mixed high-energy radiation fields

Alanine and Radio-Photo-Luminescence (RPL) dosemeters are passive dosemeters used to monitor absorbed dose in all kind of radiation fields. However, up to now both dosemeter types are calibrated to photon sources only. In order to study the response of RPL and alanine dosemeters to mixed high-energy particle fields like those occurring at CERN's accelerators, an irradiation campaign at the CERN-EC High-Energy Reference field Facility (CERF-field) was performed. In this facility a copper target is irradiated by hadrons with a momentum of 120 GeV/c. Dosemeters were exposed to various mixed radiation fields by placing them at various positions on the surface of the target. In addition to the experiment FLUKA Monte Carlo simulations were carried out, which provide information concerning the energy deposition at the dosemeter locations. This paper compares the measurements with the simulation results and discusses the radiation field compositions present at the various dosemeter positions on the target.     

Extremity ring dosimetry intercomparison in reference and workplace fields

An intercomparison of ring dosemeters has been organised with the aim of assessing the technical capabilities of available extremity dosemeters and focusing on their performance at clinical workplaces with potentially high extremity doses. Twenty-four services from 16 countries participated in the intercomparison. The dosemeters were exposed to reference photon ((137)Cs) and beta ((147)Pm, (85)Kr and (90)Sr/(90)Y) fields together with fields representing realistic exposure situations in interventional radiology (direct and scattered radiation) and nuclear medicine ((99 m)Tc and (18)F). It has been found that most dosemeters provided satisfactory measurements of H(p)(0.07) for photon radiation, both in reference and realistic fields. However, only four dosemeters fulfilled the established requirements for all radiation qualities. The main difficulties were found for the measurement of low-energy beta radiation. Finally, the results also showed a general under-response of detectors to (18)F, which was attributed to the difficulties of the dosimetric systems to measure the positron contribution to the dose.     

Radioprotection calculations for MEGAPIE

The MEGAwatt PIlot Experiment (MEGAPIE) liquid lead-bismuth spallation neutron source will commence operation in 2006 at the SINQ facility of the Paul Scherrer Institut. Such an innovative system presents radioprotection concerns peculiar to a liquid spallation target. Several radioprotection issues have been addressed and studied by means of the Monte Carlo transport code, FLUKA. The dose rates in the room above the target, where personnel access may be needed at times, from the activated lead-bismuth and from the volatile species produced were calculated. Results indicate that the dose rate level is of the order of 40 mSv h(-1) 2 h after shutdown, but it can be reduced below the mSv h(-1) level with slight modifications to the shielding. Neutron spectra and dose rates from neutron transport, of interest for possible damage to radiation sensitive components, have also been calculated.