M. Boschung

Measurements of the neutron dose equivalent for various radiation qualities, treatment machines and delivery techniques in radiation therapy

In radiation therapy, high energy photon and proton beams cause the production of secondary neutrons. This leads to an unwanted dose contribution, which can be considerable for tissues outside of the target volume regarding the long term health of cancer patients. Due to the high biological effectiveness of neutrons in regards to cancer induction, small neutron doses can be important. This study quantified the neutron doses for different radiation therapy modalities. Most of the reports in the literature used neutron dose measurements free in air or on the surface of phantoms to estimate the amount of neutron dose to the patient. In this study, dose measurements were performed in terms of neutron dose equivalent inside an anthropomorphic phantom. The neutron dose equivalent was determined using track etch detectors as a function of the distance to the isocenter, as well as for radiation sensitive organs. The dose distributions were compared with respect to treatment techniques (3D-conformal, volumetric modulated arc therapy and intensity-modulated radiation therapy for photons; spot scanning and passive scattering for protons), therapy machines (Varian, Elekta and Siemens linear accelerators) and radiation quality (photons and protons). The neutron dose equivalent varied between 0.002 and 3 mSv per treatment gray over all measurements. Only small differences were found when comparing treatment techniques, but substantial differences were observed between the linear accelerator models. The neutron dose equivalent for proton therapy was higher than for photons in general and in particular for double-scattered protons. The overall neutron dose equivalent measured in this study was an order of magnitude lower than the stray dose of a treatment using 6 MV photons, suggesting that the contribution of the secondary neutron dose equivalent to the integral dose of a radiotherapy patient is small.

Radiation doses from phantom measurements at high-pitch dual-source computed tomography coronary angiography

OBJECTIVE To compare radiation doses delivered at prospectively ECG-triggered sequential- (SEQ), retrospectively ECG-gated spiral- (RETRO) and prospectively ECG-triggered high-pitch spiral- (HP) computed tomography coronary angiography (CTCA) protocols, as well as catheter coronary angiography (CCA) using an anthropomorphic phantom. MATERIALS AND METHODS An anthropomorphic Alderson phantom equipped with 50 thermoluminescent dosimeters (TLDs) was scanned using different CTCA protocols and an uncomplicated diagnostic CCA examination was simulated. Absorbed doses were experimentally determined and effective doses calculated using the dose-length product (DLP) for CTCA and the dose-area product (DAP) for CCA, as well as according to International Commission on Radiation Protection (ICRP) publications 60 and 103. RESULTS Effective organ doses were significantly lower for HP protocols (100kV: 0.17±0.26mSv; 120kV: 0.26±0.39mSv) compared to SEQ protocols (100kV: 0.50±0.79mSv; 120kV: 0.90±1.41mSv; each p<0.05) and compared to RETRO protocols (100kV: 1.59±2.12mSv; 120kV: 2.75±3.50mSv; each p<0.05). Effective organ doses at HP-CTCA tended to be lower than at CCA (0.37±0.40mSv), however this was not statistically significant (p=0.13). Effective doses calculated according to ICRP guidelines could be estimated using the DLP and a conversion coefficient of k=0.034mSv/[mGycm] (ICRP103) or k=0.028mSv/[mGycm] (ICRP60), respectively. HP-CTCA led to a dose reduction of 89% compared to RETRO-CTCA, regardless of the calculation method used. CONCLUSIONS Radiation doses as determined by phantom measurements are significantly lower at HP-CTCA compared to SEQ-CTCA and RETRO-CTCA and comparable to uncomplicated diagnostic CCA.

DETERMINATION OF THE RESPONSE FUNCTION FOR TWO PERSONAL NEUTRON DOSEMETER DESIGNS BASED ON PADC

Since 1998 neutron dosimetry based on PADC (poly allyl diglycol carbonate) is done with a so-called original Paul Scherrer Institute (PSI) design at PSI. The original design (i.e. holder) was later changed. Both designs are optimised for use in workplaces around high-energy accelerators, where the neutron energy spectra are dominated by fast neutrons ranging up to some 100 MeV. In addition to the change of the dosemeter design a new evaluation method based on a microscope scanning technique has been introduced and the etching conditions have been optimised. In the present work, the responses obtained with the original and the new dosemeter designs are compared for fields of radionuclide sources and monoenergetic reference fields using the new evaluation method. The response curves in terms of the personal dose equivalent for normally incident neutrons were built as functions of the incident neutron energy.

Characterisation of the PSI whole body counter by radiographic imaging

A joint project between the Paul Scherrer Institut (PSI) and the Institute of Radiation Physics was initiated to characterise the PSI whole body counter in detail through measurements and Monte Carlo simulation. Accurate knowledge of the detector geometry is essential for reliable simulations of human body phantoms filled with known activity concentrations. Unfortunately, the technical drawings provided by the manufacturer are often not detailed enough and sometimes the specifications do not agree with the actual set-up. Therefore, the exact detector geometry and the position of the detector crystal inside the housing were determined through radiographic images. X-rays were used to analyse the structure of the detector, and (60)Co radiography was employed to measure the core of the germanium crystal. Moreover, the precise axial alignment of the detector within its housing was determined through a series of radiographic images with different incident angles. The hence obtained information enables us to optimise the Monte Carlo geometry model and to perform much more accurate and reliable simulations.

COMPARISON OF DIFFERENT PADC MATERIALS AND ETCHING CONDITIONS FOR FAST NEUTRON DOSIMETRY

Etched-track polyallyl diglycol carbonate (PADC) dosemeters have been in use at the Paul Scherrer Institute since 1998 in neutron dosimetry for individual monitoring. In the last years, the availability of PADC materials from different manufacturers has grown, and different etching conditions were proposed, with the intention to improve the quality and overall performance of PADC in individual neutron monitoring. The goal of the present study was to compare the performance of different PADC materials and to investigate the influence of different etching conditions on sensitivity to fast neutrons and lower detection limit. The comparison covers six different PADC materials and eight different etching conditions.

Inter-comparison among fast neutron dosimetric services using PADC material of different composition

Abstract A trial inter-comparison has been performed among four fast neutron dosimetric services: PSI(CH), ENEA (I), DRPS (UK), LANL (US). The PADC used for the tests has been produced by Intercast Europe S.p.A. Three sets of detectors have been employed: two of PADC standard material from two different batches, and one of PADC with the addition of 0.1% dioctylphthalate. Each set consisted of 50 detectors. For each set of detectors, 20 have been irradiated free-in-air at 1 mSv of H ∗ (10) with an 241 Am–Be source at ENEA-IRP, whilst the other detectors have been used as background samples. For each batch the value of the average background signal, B, the average neutron sensitivity, S, and minimum detectable dose equivalent, MDDE, have been determined. Two identical tests have been completed and separated with a time of 4 months in order to evaluate the ageing effect on the material stored in different conditions. Each dosimetric service processed the detectors according to local routine procedures. Three laboratories used an Autoscan60 reader, whilst one laboratory has an in-house reading system. Therefore, the results of the tests allowed a comparison of either the performance PADC materials, of different batches and of different compositions, or to evaluate how different etching, reading and storage conditions affect the results.

A comparison of the response of PADC neutron dosemeters in high-energy neutron fields

Within the framework of the EURADOS Working Group 11, a comparison of passive neutron dosemeters in high-energy neutron fields was organised in 2011. The aim of the exercise was to evaluate the response of poly-allyl-glycol-carbonate neutron dosemeters from various European dosimetry laboratories to high-energy neutron fields. Irradiations were performed at the iThemba LABS facility in South Africa with neutrons having energies up to 66 and 100 MeV.

Comparison of different PADC materials for neutron dosimetry

Investigations on track density and track size distributions of different PADC (poly allyl diglycol carbonate) materials have been performed. The PADC used for the tests has been produced by Thermo Electron (USA), Track Analysis System Limited (UK), Chiyoda Technol Corporation (Japan) and Intercast srl (Italy). For each PADC material 120 detectors were randomly selected out of 2 sheets: 60 detectors from one sheet have been irradiated with a personal dose equivalent of 3 mSv in the field of a (241)Am-Be source at the calibration laboratory of PSI, whilst the other 60 detectors from the other sheet have been used as background samples. All detectors have been processed according to an identical etching procedure and have been analysed with TASLImage scanning system. For each set of detectors the value of the average background signal, the average neutron sensitivity and the detection limit with respect to a personal dose equivalent measured with a dosemeter based on PADC have been determined. The results of the investigations allowed a comparison of the neutron sensitivity and background signal behaviours of PADC materials from different manufacturers and the assessment of the variation of neutron sensitivity and background signal over a single sheet.

Monitor units are not predictive of neutron dose for high-energy IMRT.

BackgroundDue to the substantial increase in beam-on time of high energy intensity-modulated radiotherapy (>10 MV) techniques to deliver the same target dose compared to conventional treatment techniques, an increased dose of scatter radiation, including neutrons, is delivered to the patient. As a consequence, an increase in second malignancies may be expected in the future with the application of intensity-modulated radiotherapy. It is commonly assumed that the neutron dose equivalent scales with the number of monitor units.MethodsMeasurements of neutron dose equivalent were performed for an open and an intensity-modulated field at four positions: inside and outside of the treatment field at 0.2 cm and 15 cm depth, respectively.ResultsIt was shown that the neutron dose equivalent, which a patient receives during an intensity-modulated radiotherapy treatment, does not scale with the ratio of applied monitor units relative to an open field irradiation. Outside the treatment volume at larger depth 35% less neutron dose equivalent is delivered than expected.ConclusionsThe predicted increase of second cancer induction rates from intensity-modulated treatment techniques can be overestimated when the neutron dose is simply scaled with monitor units.

SUITABILITY OF PORTABLE RADIONUCLIDE IDENTIFIERS FOR EMERGENCY INCORPORATION MONITORING

The suitability of portable nuclide inspectors for incorporation measurements were tested with three probes (LaBr3(Ce), NaI(Tl) and HPGe) differing in sensitive volume and energy resolution. The efficiencies for the measurement of whole-body and lung radionuclide burden were calibrated using a whole-body block phantom with traceable radionuclide sources of 60Co, 133Ba, 137Cs, 152Eu and 40K. A standing geometry was chosen as it allows rapid positioning of persons for the measurements. Decision and detection limits were determined for the unshielded detector in a normal laboratory radiation environment according to ISO 11929 for 134Cs, 137Cs and 60Co. The detection limits of all three probes were significantly higher compared to well-shielded dedicated whole-body monitors (HPGe and NaI(Tl)) using a sitting geometry. Nevertheless, lung and whole-body burdens derived from dose constraints for routine and emergency conditions could be measured with all three probes with a counting time of one minute.