Dosimetric audits of photon beams in radiation therapy centres in Rio de Janeiro, Brazil

Data related to 11 y of high-energy photon radiotherapy beam dosimetry are presented and analysed. Dosimetric evaluations were carried out using water phantoms and thimble ionisation chambers and are part of the radiation protection regulatory licensing process for medicine facilities of Brazilian government. Measurements were done at reference conditions for a standard absorbed dose of 100 cGy [cGy (=1 rad)]. The absolute per cent deviation between the measured and presumed delivered doses should not exceed the tolerance level of +/-3%. The first dosimetry survey from 1996 to 1998 showed a situation that was an object of concern. Deviations of 22 and 18.7% could be measured, although small deviations were also obtained. After 1998, the improvement in dosimetry quality control by the radiotherapy centres became clear, with most of the deviations situated within the +/-3% range. The decrease in the measured deviations presents the effective success of the Institute of Radiation Protection and Dosimetry audit programme for the improvement in the control of radiotherapy photon beams in Rio de Janeiro. Also, it is possible to recommend to Brazilian regulatory organisation a decrease in the tolerance level for dosimetric deviations in order to achieve a more precise dose delivered to patients in radiotherapy centres.     

Comparison of effective doses from various monoenergetic particles based on the stylised and the VIP-Man tomographic models

This study compares the effective doses from a MIRD-type stylised model with those derived from the scaled-down version of the tomographic VIP-Man model for photon, electron, neutron and proton beams. The effective dose results from these two models show that they differ from each other within approximately 10% for common high-energy photon beams, within approximately 16% for neutrons, and within approximately 4% for high-energy proton beams. However, for low-energy protons and common electron beams, the effective doses can be different in >100%. It is concluded that the use of a single tomographic models will not improve the operational radiation protection dosimetry involving external beam exposures.     

Application of TEPC microdosimetry to boron neutron capture therapy

Boron neutron capture therapy (BNCT) is a bimodal radiation therapy used primarily for highly malignant gliomas. Tissue-equivalent proportional counter (TEPC) microdosimetry has proven an ideal dosimetry technique for BNCT, facilitating accurate separation of the photon and neutron absorbed dose components, assessment of radiation quality and measurement of the BNC dose. A miniature dual-TEPC system has been constructed to facilitate microdosimetry measurements with excellent spatial resolution in high-flux clinical neutron capture therapy beams. A 10B-loaded TEPC allows direct measurement of the secondary charged particle spectrum resulting from the BNC reaction. A matching TEPC fabricated from brain-tissue-equivalent plastic allows evaluation of secondary charged particle spectra from photon and neutron interactions in normal brain tissue. Microdosimetric measurements performed in clinical BNCT beams using these novel miniature TEPCs are presented, and the advantages of this technique for such applications are discussed.     

Dosimetry with a transportable water calorimeter in neutron, proton and heavy-ion radiation fields

A compact and transportable water calorimeter has been developed and extensively tested in the intensive, collimated neutron field of the PTB. It has been applied for absorbed dose to water measurements in the neutron therapy field of the University of Essen, in the proton therapy fields of the HMI in Berlin and at the iThemba therapy centre near Cape Town, South Africa, as well as in the (12)C-beam of the therapy facility at GSI in Darmstadt, Germany. Absolute dosimetry with relative standard uncertainties of less than 1.8% was achieved in all radiation fields. The results obtained using the water calorimeter are compared with the ionisation chamber measurements in the same radiation fields. The heat defect for the water in the calorimeter core was determined separately in independent measurements by irradiation with different charged particle beams covering a wide range of linear energy transfer.     

A toolkit for epithermal neutron beam characterisation in BNCT

Methods for dosimetry of epithermal neutron beams used in boron neutron capture therapy (BNCT) have been developed and utilised within the Finnish BNCT project as well as within a European project for a code of practise for the dosimetry of BNCT. One outcome has been a travelling toolkit for BNCT dosimetry. It consists of activation detectors and ionisation chambers. The free-beam neutron spectrum is measured with a set of activation foils of different isotopes irradiated both in a Cd-capsule and without it. Neutron flux (thermal and epithermal) distribution in phantoms is measured using activation of Mn and Au foils, and Cu wire. Ionisation chamber (IC) measurements are performed both in-free-beam and in-phantom for determination of the neutron and gamma dose components. This toolkit has also been used at other BNCT facilities in Europe, the USA, Argentina and Japan.     

Studies on the response of the TLD badge for high-energy photons

Absorbed tissue dose measurements are carried out for high-energy photon beams using CaSO4:Dy thermo-luminescence dosemeter (TLD) badge and the results are also verified using ionisation chamber used in radiation therapy. The photon beams generated using linear accelerator at 6 and 18 MV photon beam energies have been used and the absorbed doses are measured at the surface as well as at various depths. It has been found that the depth at which maximum dose is delivered increases with the increase in photon energy and the depth of maximum absorbed dose in tissue occurs beyond 10 mm. It has also been found that the evaluation of the absorbed dose (or Hp(10) as well) using thermoluminescence readout of disc D1 clearly shows that the current TLD badge provides a reasonable estimate of the effective dose for photon fields from 6 to 18 MV linacs for anterior-posterior incidence. The paper also provides information regarding the misinterpretation of radiation pattern in multi-element/filter TLD badge.     

医用质子加速器输出剂量的测量

本文较详细叙述了国际原子能机构的(IAEA)第398号报告中质子外照射辐射治疗源吸收剂量的测量,并介绍国外质子治疗中心的临床和设备的有关规范指标。     

AAPM TG-51临床参考剂量学的特点及应用

本文介绍了AAPM (美国医学物理家协会 )TG - 5 1号新报告 ,即“高能光子和电子束参考剂量学议定书”。它是以60 Coγ射线在水中的吸收剂量直接校准治疗水平电离室型剂量计 ,然后根据临床应用的射束辐射质再计算水中的吸收剂量。它是辐射剂量学中的一项重大改进 ,这种方法易于理解 ,执行方便 ,精确度高。     

Ion chamber gas-to-wall conversion factors for fast neutron dosimetry

Modern ionising photon dosimetry is essentially entirely based upon gas-filled cavity determinations. For photons, ion chamber response is largely independent of photon energy almost perfectly transforming absorbed dose in the gas to the surrounding media. Absolute uncertainties are <1-2%. For fast neutron dosimetry, this is certainly not the case. Interpretation of the response of the cavity filling material, usually a gas, to the charged particle spectrum induced in the walls and interacting with the cavity gas is fraught with uncertainties. Despite these challenges, gas filled cavities surrounded by various mixtures, compounds and elements, have proved to be essential for integral determinations of the indirectly ionising neutrons, generating dosimetric quantities, such as kerma and absorbed dose. The transformation from gas response to wall dose is material dependent and varies with neutron energy. This study discusses recent advances in cavity response interpretation using the results from complex nuclear modelling of microscopic cross sections as well as estimates of secondary particle production enabling much improved cavity gas-to-wall media conversion factors.     

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.     

Absorbed dose measurements of a handheld 50 kVP X-ray source in water with thermoluminescence dosemeters

Absorbed dose rate measurements of a 50 kV(p) handheld X-ray probe source in a water phantom are described. The X-ray generator is capable of currents of up to 40 microA, and is designed for cranial brachytherapy and intraoperative applications with applicators. The measurements were performed in a computer-controlled water phantom in which both the source and the detectors are mounted. Two different LiF thermoluminescence dosemeter (TLD) phosphors were employed for the measurements, MTS-N (LiF:Mg,Ti) and MCP-N (LiF:Mg,Cu,P). Two small ionisation chambers (0.02 and 0.0053 cm(3)) were also employed. The TLDs and chambers were positioned in watertight mounts made of water-equivalent plastic. The chambers were calibrated in terms of air-kerma rate, and conventional protocols were used to convert the measurements to absorbed dose rate. The TLDs were calibrated at National Institute of Standards and Technology (NIST) in terms of absorbed dose rate using a (60)Co teletherapy beam and narrow-spectrum X-ray beams. For the latter, absorbed dose was inferred from air-kerma rate using calculated air-kerma-to-dose conversion factors. The reference points of the various detectors were taken as the center of the TLD volumes and the entrance windows of the ionisation chambers. Measurements were made at distances of 3-45 mm from the detector reference point to the source center. In addition, energy dependence of response measurements of the TLDs used was made using NIST reference narrow spectrum X-ray beams. Measurement results showed reasonable agreement in absorbed dose rate determined from the energy dependence corrected TLD readings and from the ionisation chambers. Volume averaging effects of the TLDs at very close distances to the source were also evident.     

Neutron production in tissue-like media and shielding materials irradiated with high-energy ion beams

Secondary neutrons produced in high-energy therapeutic ion beams require special attention since they contribute to the dose delivered to patient, both to tumour and to the healthy tissues. Moreover, monitoring of neutron production in the beam line elements and the patient is of importance for radiation protection aspects around ion therapy facility. Monte Carlo simulations of light ion transport in the tissue-like media (water, A-150, PMMA) and materials of interest for shielding devices (graphite, steel and Pb) were performed using the SHIELD-HIT and MCNPX codes. The capability of the codes to reproduce the experimental data on neutron spectra differential both in energy and angle is demonstrated for neutron yield from the thick targets. Both codes show satisfactory agreement with the experimental data. The absorbed dose due to neutrons produced in the water and A-150 phantoms is calculated for proton (200 MeV) and carbon (390 MeV/u) beams. Secondary neutron dose contribution is approximately 0.6% of the total dose delivered to the phantoms by proton beam and at the similar level for both materials. For carbon beam the neutron dose contribution is approximately 1.0 and 1.2% for the water and A-150 phantoms, respectively. The neutron ambient dose equivalent, H(10), was determined for neutrons leaving different shielding materials after irradiation with ions of various energies.     

Intercomparison of absorbed dose to water measurements for 60Co gamma rays using Fricke,alanine and radiochromic dye film dosimetry

Measurements of absorbed dose at 5 cm depth in a 30 x 30 x 30 cm3 water phantom have been performed using three independent dosimetric techniques: Fricke, alanine and radiochromic dye film (GafChromic HD-810). The measurements were carried out in the secondary standard dosimetry laboratory at ININ Mexico using a collimated 60Co gamma source with a radiation field of 10 x 10 cm2 at the phantom front surface. The source to phantom distance was set at 100 cm. The reference absorbed dose at 5 cm depth in the water phantom was obtained using a 0.6 cm3 ionisation chamber. The absorbed dose to water for the test dosimetry techniques was around 100 Gy. The deviations of the dose obtained from these dosimetry techniques were within 4%. The reasons for these deviations are discussed.     

Dosimetric measurements with a brain equivalent plastic walled ionization chamber in an epithermal neutron beam

The tissue substitute A-181 plastic, which has an elemental composition matching both the constituent hydrogen and nitrogen of brain tissue, was assessed for dosimetry in boron neutron capture therapy (BNCT). The sensitivity of an A-181 walled ionization chamber relative to photons for all neutrons in a clinical epithermal beam was calculated to vary between 0.79 +/- 0.04 in-air and 0.95 +/- 0.01 at depths of 4 cm and greater in-phantom. Differences in the total neutron doses measured with A-150 and A-181 plastic-walled chambers were attributed, within experimental error, to the dose produced by thermal neutron capture reactions from the different concentrations of nitrogen in the two tissue substitutes. The response of the A-181 chamber was converted to total neutron dose with an uncertainty increasing with depth in-phantom from 13 to 23% the magnitude of which is determined by the subtraction of a relatively large photon dose. The use of A-181 in place of A-150 plastic will no longer require partitioning the measured neutron dose by energy and should simplify dose reporting in BNCT.     

Study of materials for mixed field dosimetry by EPR spectroscopy

EPR dose reconstruction after accidental photon exposure based on materials irradiated in the vicinity of the victim (sucrose, medicine tablets, etc.) was used successfully in several cases referenced in the literature. However, accidental exposure may also occur with a neutron component such as in the Tokai-Mura criticality accident. The aim of this work is to investigate the potentiality of EPR dosimetry for mixed photon and neutron field exposure with different organic materials already used for photon exposure (sucrose) or with potential dosimetric properties (ascorbic acid, sorbitol, glucose, galactose, fructose, lactose and mannose). To assess the neutron sensitivity, the materials were exposed to a mixed radiation field of an experimental reactor with different neutron to photon ratios. The relative neutron sensitivity was found to range from 12 to 43% according to the materials. The potentiality of these materials for mixed field EPR dosimetry is discussed.     

Investigations of recombination chambers for BNCT beam dosimetry

A set of cylindrical recombination chambers, including a tissue-equivalent chamber and three graphite chambers filled with different gases-CO(2), N(2) and (10)BF(3), was designed for the dosimetry of therapeutic neutron radiation beams used for BNCT. The separation of the dose components is based on differences of the shape of the saturation curve depending on the LET spectrum of the investigated radiation. The measurements using all the chambers were performed in a reactor beam of NRI ReZ (Czech Republic) and in the reference radiation fields of a (252)Cf radiation source free in air or in filters.     

An accelerator-based epithermal neutron beam design for BNCT and dosimetric evaluation using a voxel head phantom

The beam shaping assembly design has been investigated in order to improve the epithermal neutron beam for accelerator-based boron neutron capture therapy in intensity and quality, and dosimetric evaluation for the beams has been performed using both mathematical and voxel head phantoms with MCNP runs. The neutron source was assumed to be produced from a conventional 2.5 MeV proton accelerator with a thick (7)Li target. The results indicate that it is possible to enhance epithermal neutron flux remarkably as well as to embody a good spectrum shaping to epithermal neutrons only with the proper combination of moderator and reflector. It is also found that a larger number of thermal neutrons can reach deeply into the brain and, therefore, can reduce considerably the treatment time for brain tumours. Consequently, the epithermal neutron beams designed in this study can treat more effectively deep-seated brain tumours.     

Dosimetry at 28 keV synchrotron radiation for cell irradiations

Accurate dosimetry is a prerequisite for reliable comparisons between radiobiological irradiation experiments. Parameters affecting the determination of absorbed dose to cells in the shape of a small cell pellet in a centrifuge tube, irradiated by 28 keV mono-energetic photons from a synchrotron, were investigated. Thermoluminescent dosimeter (TLD), diode and ion chambers were utilized to monitor the irradiations. The distribution of the absorbed dose and such parameters as scatter, attenuation and interface dosimetry in the target, which influence the dose, were studied. A method for inter-calibrations of two different calibration sources by using TLD and TLD readers is given. Characteristics of the TLD, that is, fading, supralinearity, energy response, self-attenuation and mini-dosimetry were considered for the dosimetry. A method for correcting photon fluence attenuation in cylindrical TLDs is presented. The study shows that the absorbed dose to cells irradiated at low photon energy at a synchrotron irradiation facility can, using accurate dosimetry protocol, be correctly and reproducibly determined.     

Patient neutron dose equivalent exposures outside of the proton therapy treatment field

A large fraction of dose to healthy tissue located outside of the treatment field during proton therapy is attributable to neutrons produced in the beam-delivery apparatus. In this work, the neutron dose equivalent (H) per therapeutic proton absorbed dose (D) was estimated for typical treatment conditions as a function of range modulation width, angle with respect to the incident proton beam, and the distance from the isocentre at the Harvard Cyclotron Laboratory's (Cambridge, MA) passively spread treatment field using Monte Carlo simulations. For a beam with 16 cm penetration (depth) and a 5 x 5 cm2 lateral field size at the patient location along the incident beam direction at 100 cm from the isocentre, the predicted H/D values are 0.35 and 0.60 mSv Gy(-1) from the simulations and measurements, respectively. At all locations, the predicted H/D values are within a factor of 2 and 3 of the measured result for no modulation and 8.2 cm of modulation, respectively.     

Analytical shielding calculations for a proton therapy facility

The University of Pennsylvania is building a proton therapy facility in collaboration with Walter Reed Army Medical Center. The proposed facility has four gantry rooms, a fixed beam room and a research room, and will use a cyclotron as the source of protons. In this study, neutron shielding considerations for the proposed proton therapy facility were investigated using analytical techniques and Monte Carlo simulated neutron spectra. Neutron spectra calculations were done using the GEANT4 (v6.2) simulation code for various materials: water, carbon, iron, nickel and tantalum to estimate the neutron production at proton beam energies of 100, 175 and 250 MeV. Dose equivalent calculations were performed using analytical methods at various critical points within the facility, by folding the GEANT4 produced neutron spectra with dose equivalent rate data from the National Council on Radiation Protection and Measurements (NCRP) Report #144.