Use of water-equivalent plastic scintillator for intravascular brachytherapy dosimetry

Beta irradiation has recently been investigated as a possible technique for the prevention of restenosis in intravascular brachytherapy after balloon dilatation or stent implantation. Present methods of beta radiation dosimetry are primarily conducted using radiochromic film. These film dosimeters exhibit limited sensitivity and their characteristics differ from those of tissue, therefore the dose measurement readings require correction factors to be applied. In this work a novel, mini-size (2 mm diameter by 5 mm long) dosimeter element fabricated from Organic Plastic Scintillator (OPS) material was employed. Scintillation photon detection is accomplished using a precision photodiode and innovative signal amplification and processing techniques, rather than traditional photomultiplier tube methods. A significant improvement in signal to noise ratio, dynamic range and stability is achieved using this set-up. In addition, use of the non-saturating organic plastic scintillator material as the detector enables the dosimeter to measure beta radiation at very close distances to the source. In this work the plastic scintillators have been used to measure beta radiation dose at distances of less than 1 mm from an Sr-90 cardiovascular brachytherapy source having an activity of about 2.1 GBq beta radiation levels for both depth-distance and longitudinal profile of the source pellet chain, both in air and in liquid water, are measured using this system. The data obtained is compared with results from Monte Carlo simulation technique (MCNP 4B). Plastic scintillator dosimeter elements, when used in conjunction with photodiode detectors may prove to be useful dosimeters for cardiovascular brachytherapy beta sources, or other applications where precise near-source field dosimetry is required. The system described is particularly useful where measurement of actual dose rate in real time, a high level of stability and repeatability, portability, and immediate access to results are prime requirements.     

Characterization of a new commercial single crystal diamond detector for photon- and proton-beam dosimetry

A synthetic single crystal diamond detector (SCDD) is commercially available and is characterized for radiation dosimetry in various radiation beams in this study. The characteristics of the commercial SCDD model 60019 (PTW) with 6- and 15-MV photon beams, and 208-MeV proton beams, were investigated and compared with the pre-characterized detectors: Semiflex (model 31010) and PinPoint (model 31006) ionization chambers (PTW), the EDGE diode detector (Sun Nuclear Corp) and the SFD Stereotactic Dosimetry Diode Detector (IBA). To evaluate the effects of the pre-irradiation, the diamond detector, which had not been irradiated on the day, was set up in the water tank, and the response to 100 MU was measured every 20 s. The depth–dose and profiles data were collected for various field sizes and depths. For all radiation types and field sizes, the depth–dose data of the diamond chamber showed identical curves to those of the ionization chambers. The profile of the diamond detector was very similar to those of the EDGE and SFD detectors, although the Semiflex and PinPoint chambers showed volume-averaging effects in the penumbrae region. The temperature dependency was within 0.7% in the range of 4–41°C. A dose of 900 cGy and 1200 cGy was needed to stabilize the chamber to the level within 0.5% and 0.2%, respectively. The PTW type 60019 SCDD detector showed suitable characteristics for radiation dosimetry, for relative dose, depth–dose and profile measurements for a wide range of field sizes. However, at least 1000 cGy of pre-irradiation will be needed for accurate measurements.     

Reconstruction of external dose from beta radiation sources of nuclear weapon origin

In response to requests from the Department of Veterans Affairs, a methodology was developed to assess the external dose accrued by personnel in the vicinity of beta radiation sources of nuclear weapon origin. This methodology has been applied in support of the Nuclear Test Personnel Review (NTPR) Program implemented by the Department of Defense. As required by the Code of Federal Regulations (Title 32, Part 218 and Title 38, Part 3.311), the NTPR Program must evaluate radiological hazards from nuclear testing activities that include alpha particle, beta particle, neutron, and photon emissions from radionuclides. Prior to the development of this methodology, only photon and neutron radiations were explicitly quantified for external dose assessments in this program. Alpha radiation of external origin presents no risk for biological effects due to external dose potential to skin tissue because of the particle's very limited range. However, beta particles are sufficiently penetrating to have such potential. Methods are provided to quantify ionizing radiation doses to the skin and lens of the eye from beta radiation sources of nuclear weapon origin located external to the body. This formulation allows the estimation of beta dose from a film badge (gamma) dose or from an equivalent reconstructed gamma dose.     

A small, multi-element fingertip dosimeter for beta radiation

Many occupations involve potential exposures, directly or indirectly, to sources of beta radiation. The region of highest exposure, in many cases, will be the extremities and in particular the fingertips of people handling beta sources. Because current extremity dosimeters can significantly underestimate beta doses to the fingertips, an improved fingertip beta dosimeter was developed. The dosimeter employs a multi-element, multi-filter concept by stacking four or five thin (0.13 mm) LiF-Teflon TLDs to form the beta detector element. The entire dosimeter is approximately 1 mm thick, flexible and rugged enough for field use without interference of the user's manual dexterity. The fingertip dosimeter provides data for determining the beta energy of each exposure. The beta energy can be used to determine the TLD correction factor for converting the TLD output to beta dose. The data can be used to reconstruct the beta depth dose curve and the depth dose curve can be used to calculate the beta dose in tissue at 7 mg cm-2, for legal reporting purposes, or at any other desired depth within the range of the beta radiation. Relationships between the effective mass absorption coefficient and beta energy and between the beta correction factor and beta energy were determined for use in this study. Beta sources were fabricated for these studies, and an extrapolation chamber was used to determine reference beta doses. Tests of the fingertip dosimeter were performed by exposing it to single beta sources and to multiple beta sources. The dosimeter should be useful for monitoring exposures to beta energies ranging from 0.29-2.5 MeV.     

关于CTLD-J4000型个人剂量计用于能量鉴别的研究

目的 通过对分散性为±1%探测器计量检定结果的分析,验证CTLD-J 4000型胸章式个人剂量计能否适用于不同能量的放射诊疗场所。方法 根据《个人与环境监测用X、γ辐射热释光剂量测量(装置)系统》(JJG 593-2006)的要求,将准备的剂量计送至国防科技工业电离辐射一级计量站进行检定,分析结果。结果 热释光剂量监测系统的线性、能量响应符合检定规程要求;通过确定剂量计不同槽内探测器计数比值,能够区分不同能量的光子,从而科学选择刻度因子,计算人员有效剂量。结论 CTLD-J4000型胸章式个人剂量计能够鉴别能量,适用于在不同能量放射诊疗场所的个人剂量监测工作。     

EBT3胶片和辐射荧光屏测量质子重离子放疗设备照射野剂量分布的对比研究

目的 对比离线分析的EBT3胶片和在线实时分析的辐射荧光屏对医用质子重离子加速器照射野剂量分布的质控测量差异,选择更适合的设备质控测量的探测器系统。方法 针对主动式点扫描医用质子重离子加速器,在布喇格峰展宽为3 cm,射程分别为7、16、29 cm条件下,通过放疗计划系统规划100 mm×100 mm的质子和碳离子照射野,分别利用EBT3胶片和辐射荧光屏探测器测量各种条件下的照射野大小,EBT3胶片图像获取在照射24小时后通过平板扫描仪获取,辐射荧光屏图像在束流照射过程中由相机实时拍照获取,通过照射野剂量分布曲线的半高宽进行对比分析。结果 三种质子和碳离子射程条件下,通过对比分析,不论物理意义方面还是统计学方面,两种探测器测量质子重离子照射野剂量分布,结果一致性好,精度相当,辐射荧光屏探测器在结果处理时间上更有效率。结论 两种探测器都满足质子重离子照射野剂量学参数测量要求,在照射野剂量分布的指标测量方面,能够在线实时测量分析的辐射荧光屏探测器比EBT3胶片更符合放疗质控发展要求。     

Comparison of two different types of LiF:Mg,Cu,P thermoluminescent dosimeters for detection of beta rays (beta-TLDs) from 90Sr/90Y, 85Kr and 147Pm sources

Targeted radionuclide therapies in nuclear medicine departments increasingly depend on using unsealed beta radiation sources in the labeling of peptides and antibodies. Monitoring doses received by the fingers and hands during these procedures is best accomplished with TLD dosimeters that can be located at the fingertips. The present study examines the response of two TLD dosimeters (MCP-Ns and GR200A) to 90Sr/90Y, 85Kr, and 147Pm. The dosimeters were supplied by two different services, and all irradiations were performed at the PTB Institute in Germany. Each dosimetry service evaluated the dosimeters without knowledge that they had been purposefully irradiated. The accuracy and precision of the dosimeters were evaluated as a function of delivered dose, energy of beta particles and angular incidence. The results are compared to performance measures recommended by the IEC. Both dosimeter types displayed significant energy dependence. Angular dependence was moderate. Accuracy and precision as a function of dose (linearity) differed between the two systems, with the MCP-Ns being noticeably better than the GR200A. The superior precision makes the MCP-Ns much more useful for extremity dose measurements. The differences between these two dosimeter systems reinforce the need to evaluate a dosimeter carefully before using it in the daily work routine.     

热释光剂量测量系统检定的质量保证及刻度因子的应用

目的为保证辐射受照人员个人剂量监测数据的可靠性,探讨热释光剂量测量系统检定过程影响因素,计算线性和不同能响刻度因子,确定不同类别照射群体所使用的刻度因子。方法依据标准JJG 593-2006,设计2种方案进行系统检定。方案1:已筛选退火后的探测器片放于有机玻璃平衡板中,分别予以不同剂量的137Csγ射线照射;方案2:将探测器片放于剂量计盒中,置于与方案1相同位置,予γ射线照射和窄谱过滤X射线照射。计算线性和刻度因子。结果直接照射探测器片与将探测器片放于剂量计壳中照射刻度因子差异以及不同辐射能量响应刻度因子差异均具有统计学意义。结论热释光剂量测量系统检定时应严格控制平衡条件,带剂量计盒平衡。对于放射诊断照射群体,应用低能量刻度因子;对于放射治疗照射群体,应用γ射线照射的刻度因子。2种刻度因子的使用,使监测数据更为准确。     

Environmental dosimetry performance criteria: pilot test of proposed ANSI draft standard N13.29. American National Standards Institute

American National Standards Institute (ANSI) standards are a source of technical guidance on the application of passive radiation detectors in personnel and environmental monitoring programs. The proposed new ANSI Draft N13.29 describes performance tests for environmental dosimetry providers that are analogous to those now required in personnel dosimetry. The objective of this study was to pilot test the procedural specifications before Draft N13.29 is submitted for final balloting. The results of the performance tests are relevant to environmental surveillance programs, which may be affected if Draft N13.29 is approved. Seven environmental dosimetry providers submitted dosimeters for the pilot test, which included two phases. The first phase involved exposing dosimeters to laboratory gamma, beta, and x-ray sources at routine and accident dose levels. In the second phase, dosimeters were subjected to 90 d of simulated environmental conditions in a chamber that cycled through prescribed temperature and humidity parameters and lighting conditions. Two participants passed all categories of the laboratory test phase, and all seven passed the environmental test phase. We report here on the results of the performance tests; additional results relevant to finalizing Draft N13.29 are detailed elsewhere.     

A theoretical study of the radiological properties and water equivalence of Fricke and polymer gels used for radiation dosimetry

A method is described which determines the radiological equivalence of different materials by comparing their macroscopic photon and electron interaction parameters over the energy range of interest. This method has been applied to Fricke and polymer gels used for radiation dosimetry. Photon and electron dose distributions were calculated using a Monte Carlo technique. 6 MV photon gel depth dose curves were all within 1% of those in water over the depth range studied. 6 MeV electron beam depth dose curves were all within 1 mm of those in water. The results show that for the gels investigated, at megavoltage energies the dominant property determining water equivalence is the electron density of the gel, and therefore future research into gel manufacture for external beam radiotherapy should aim to have the electron density (electrons per unit volume) as close as possible to water. However, care needs to be taken when using gels for dosimetry of photons with incident energy < 100 keV.     

Calculations for beta dosimetry using Monte Carlo code (OREC) for electron transport in water

A Monte Carlo computer code (OREC) for calculating the detailed transport and energy deposition for primary electrons and all of their secondaries in liquid water has been investigated for use in beta-ray dosimetry. Some modifications have been made in the original code for its application to tissue and tissue-equivalent materials. The code gives reasonably good agreement with beta spectral data and depth-dose curves measured in tissue-equivalent plastics for several calibrated beta sources. The calculations permit a direct evaluation of the skin dose equivalent, i.e., the dose equivalent, Hs (0.07), at a depth of 0.070 mm in tissue. Calculations are presented for monoenergetic electrons, showing the distributions in the maximum depth of penetration and in the total pathlength traveled. Direct comparisons are made between depth-dose curves calculated for 99Tc and 147Pm plaque sources and measurements made with extrapolation chambers. The energy spectrum of beta particles emerging from a thick 99Tc plaque source also is calculated, and the angular distribution is found to be almost independent of the energy. The pulse-height spectrum in a tissue-equivalent plastic scintillator calculated for this source shows good agreement with the measured spectrum. The calculations also provide the Hs(0.07) dose equivalent for the 99Tc source, which is found to be consistent with that inferred independently from the spectral measurements. A calculated curve for converting spectrometer measurements to Hs(0.07) dose equivalent is in good agreement with a semi-empirical curve that was developed independently. It appears that calculations made with the electron transport code for water can provide useful information for beta dosimetry.     

Experimental and computational techniques for beta-particle dosimetry

This paper reports the experimental and the computational techniques that were specifically developed to provide dose response data for a new method of beta dosimetry, which is reported in an accompanying article (Sh87). The specific experimental techniques consist of setting up, calibrating and obtaining backscattering and resolution corrections for a plastic-scintillator-based beta spectrometer. The computation techniques involve (1) adapting a Monte Carlo electron transport computer code to use measured beta energy distributions as input data and (2) using the code to calculate the energy deposition of these distributions of electrons in a slab of material. The energy deposition of backscattered electrons incident on the slab is also taken into account. Codes, which were developed to calculate the energy deposited by photons in LiF, are used to derive a theoretical value for the TLD response calibration factor. This factor compares well to the experimentally derived result which was obtained by exposing TLDs to a calibrated 137Cs/137mBa photon source.     

Energy dependence corrections to MOSFET dosimetric sensitivity

Metal Oxide Semiconductor Field Effect Transistors (MOSFET’s) are dosimeters which are now frequently utilized in radiotherapy treatment applications. An improved MOSFET , clinical semiconductor dosimetry system (CSDS) which utilizes improved packaging for the MOSFET device has been studied for energy dependence of sensitivity to x-ray radiation measurement. Energy dependence from 50 kVp to 10 MV x-rays has been studied and found to vary by up to a factor of 3.2 with 75 kVp producing the highest sensitivity response. The detectors average life span in high sensitivity mode is energy related and ranges from approximately 100 Gy for 75 kVp x-rays to approximately 300 Gy at 6 MV x-ray energy. The MOSFET detector has also been studied for sensitivity variations with integrated dose history. It was found to become less sensitive to radiation with age and the magnitude of this effect is dependant on radiation energy with lower energies producing a larger sensitivity reduction with integrated dose. The reduction in sensitivity is however approximated reproducibly by a slightly non linear, second order polynomial function allowing corrections to be made to readings to account for this effect to provide more accurate dose assessments both in phantom and in-vivo.     

A technical basis for the continued use of exposure as an acceptable operational quantity in radiation protection

Within the tabulated values of the new [to U.S. Department of Energy (DOE)] radiation weighting factors, it can be seen that a doubling of the neutron factor occurs for the 0.1 to 2 MeV neutron energy range. Hence, with the effective replacement of the quality factor by these new radiation weighting factors (for the protection quantities), it has been widely understood that the new changes will most definitely impact neutron dosimetry. However, it is less well understood that the new changes could also affect photon (and beta) dosimetry, i.e., photon reference fields, instrument design, and instrument calibrations. This paper discusses the ramifications, and ultimately concludes that the use of exposure for workplace measurements complies with both current and amended DOE requirements.     

Production of neutrons from water, polyethylene, tissue equivalent material and CR-39 irradiated with 2.5-30 MeV photons

Photoneutron yields from water, polyethylene, tissue substitute and CR-39 have been calculated for the photon energy range of 2 to 30 MeV, using a previously established method and photoneutron production data on hydrogen, carbon, nitrogen and oxygen. The rarer isotopes of the constituent elements of these compounds, namely 2H, 13C, 15N, 17O and 18O, have been taken into account and neutrons are shown to be produced for photon energies above 2.2 MeV, the (gamma, n) threshold for 2H. The data are useful for estimating neutron production in materials located in the vicinity of a megavoltage radiotherapy beam. Substances such as those considered here are often used as filtration, phantom or scattering material and as components of neutron dosimetry detectors. Photoneutrons produced in such materials may need to be taken into consideration when carrying out neutron dosimetry in the presence of photons in this energy range, especially when the neutron flux is several orders of magnitude less than that of the photons.     

Dosimetry of a low-kV intra-operative X-ray source using basic analytical beam models

The low energy (30-50 kVp) beams from an intra-operative X-ray source are modelled using a basic analytical model considering just primary beam attenuation and absorption. Spatial dosimetry at such low energies is difficult due to the rapid changes in dose-rate from the radiation source. The purpose of the model was to determine the variation with distance in water of coefficients required for beam dosimetry and to validate beam measurements performed in water of high-gradient dose distributions. The model predicts a change in mean mass-energy absorption coefficient of up to 3 % over the range of clinically-relevant distances in water. Distance-dose distributions (variation in dose with distance in water) for the X-ray source were calculated with the model and found to be in agreement with measurement (at clinically-relevant distances), to within a spatial distance comparable to the dimensions and positional accuracy of the ionization chamber used, and comparable to the expected dosimetric anisotropy of the radiation source. Measured and calculated distance-doses begin to diverge at relatively large distances from the radiation source, which is where dose-rates are so low that detector signal levels are comparable with noise.     

Mayak film dosimeter response studies, part I: measurements

The Mayak Worker Dosimetry study is a joint Russian/U.S. project to evaluate doses received by workers at the Mayak Production Association facilities from 1948-1972. A key investigation in this project is the characterization of responses of the three types of film dosimeters used to monitor workers during this time period. Experimental irradiations of the dosimeters were performed in the radiation calibration laboratories at the National Research Center for Environment and Health (GSF) in Munich, Germany. The irradiations used photon sources from x-ray beams with ten different energy distributions and with Co and Cs isotopic gamma sources. Irradiations were performed with the dosimeters on phantoms and free-in-air. The dosimeters and phantoms were also positioned at varying angles to the radiation beam. The result of the experiments was a thorough characterization of the dosimeter response as a function of photon energy and as a function of angle for energy and angular ranges that cover the conditions encountered in the Mayak workplaces. The characterization data were then available for use in developing correction factors, which could be applied to worker dosimeter readings to provide a more accurate assessment of worker dose and estimates of doses to organs.     

用热释光剂量计测定计算机视频终端的电离辐射剂量

目的:测定计算机视频终端的X射线剂量水平。方法:采用高灵敏LiF(Mg,Cu,P)热释光剂量计对计算机视频终端表面及其附加防护屏(网)后的辐射剂量进行累积剂量监测。结果:计算机视频终端表面的平均照射量率为2.40×10^-7C·kg^-1/d(0.93mR/d),在防护屏(网)外的平均照射量率是0.62×10^-7C·kg^-1/d(0.24mR/d)。结论:监测结果表明,计算机产生的低能和低水平电离辐射剂量未超过有关规定标准。     

Comparison of natural and synthetic diamond X-ray detectors

Diamond detectors are particularly well suited for dosimetry applications in radiotherapy for reasons including near-tissue equivalence and high-spatial resolution resulting from small sensitive volumes. However, these detectors have not become commonplace due to high cost and poor availability arising from the need for high-quality diamond. We have fabricated relatively cheap detectors from commercially-available synthetic diamond fabricated using chemical vapour deposition. Here, we present a comparison of one of these detectors with the only commercially-available diamond-based detector (which uses a natural diamond crystal). Parameters such as the energy dependence and linearity of charge with dose were investigated at orthovoltage energies (50–250 kV), and dose-rate dependence of charge at linear accelerator energy (6 MV). The energy dependence of a synthetic diamond detector was similar to that of the natural diamond detector, albeit with slightly less variation across the energy range. Both detectors displayed a linear response with dose (at 100 kV) over the limited dose range used. The sensitivity of the synthetic diamond detector was 302 nC/Gy, compared to 294 nC/Gy measured for the natural diamond detector; however, this was obtained with a bias of 246.50 V compared to a bias of 61.75 V used for the natural diamond detector. The natural diamond detector exhibited a greater dependency on dose-rate than the synthetic diamond detector. Overall, the synthetic diamond detector performed well in comparison to the natural diamond detector.