Neutron spectra and dosimetric features around an 18 mv linac accelerator

Using the difference between responses to neutrons of TLD-600 and TLD-700, three experimental devices were constructed and arranged to measure thermal neutron fluences, neutron spectra, and neutron doses inside the treatment room of a radiotherapy 18 MV Linear electron accelerator (Linac). Thermal neutron fluences were measured with TLD-600/TLD-700 pairs arranged in both a bare and a cadmium (Cd) foil covered methacrylate box. Neutron spectra were measured in 26 energy bins by introducing pairs of TLD-600/TLD-700 in air and into the middle of five polyethylene spheres with diameters of 3, 5, 8, 10, and 12 inches. A PC version of the BUNKI code was used to unfold the six measurements in each sphere to obtain the 26 energy bins. Neutron and photon doses were measured by introducing pairs of TLD-600/TLD-700 into the middle of a single 25-cm-diameter paraffin sphere. The three required neutron calibrations were carried out at the Nuclear Technology Laboratory of the Polytechnique University of Madrid (UPM), using an 241Am-Be neutron source with an alpha activity of 111 GBq and a yield of 6.6 x 10(6) neutrons s(-1). Three devices were needed for the necessary calibrations: a BF3 counter for the thermal neutron fluence calibration, a LUDLUM 42-5 Bonner spectrometer with five 0.95 g cm(-3) polyethylene spheres with a LiI(Eu) 4 x 4 mm2 scintillation counter for the neutron spectrometer calibration and a NEMO 9140 remmeter for the paraffin remmeter calibration. The Monte Carlo code MCNP 4C has been used in two ways: to calculate the neutron kerma contribution to two TLDs (type 600 and 700) both in air and inside the paraffin sphere, and to determine the neutron spectra at those Linac room zones where the neutron spectra were measured. Thermal neutron fluences of 2.9 x 10(4) +/- 8.6 x 10(3) cm(-2) s(-1), measured around the Linac head plane, and 2.3 x 10(4) +/- 2.3 x 10(3) cm(-2) s(-1), measured at the patient couch plane, are in agreement with previous independent measurements from other authors. The calculated and measured neutron spectra obtained in the treatment room showed three distinct regions: a peak around 0.1 MeV, a flat epithermal region and a thermal region with values similar to those mentioned above. Patient dose equivalents of 0.5 mSv and 5 mSv from neutrons and photons, respectively, were obtained per treatment Gray.     

Thermal neutron fluence measurement in a research reactor using thermoluminescence dosimeter TLD-600.

A thermal neutron fluence in the range between 10(11) and 10(13) n cm(-2) in the reactor core of the Tehran research reactor has been measured using TLD-600 thermoluminescence dosimeters. After a thermal treatment of 1 h at 400 degrees C followed by 20 h cooling down to room temperature of pre-exposed dosimeters in the reactor, the accumulated TL light was measured after periods of storage of 24, 48 and 72 h. The influence of the irradiation-induced damage effect on the response of TLDs and their subsequent readings has been minimized in this manner. The induced TL light due to self-activity in the TLD-600 dosimeters, which is dependent on the neutron fluence, caused a conveniently measurable TL glow curve. The induced TL in the dosimeter due to the Q-value for the beta-decay of tritium Ebeta-max = 18.6 keV has been reproduced separately by a beta source to check the proportions of radionuclides in the chip. A short theoretical treatment is also presented.     

Accelerator room photoneutron and photon background measurements using thermoluminescent dosimeters

Photoneutron dose equivalents and photon doses in the treatment room of a clinical linear accelerator were measured with sets of isotopically enriched LiF thermoluminescent dosimeters and a moderating sphere. Dosimeter neutron calibrations with 252Cf sources were repeated many times during the extended series of measurements because the 6LiF dosimeter sensitivity increased with successive neutron irradiations. Expressed as a fraction of the primary bremsstrahlung beam dose at maximum, the photoneutron background was 2.04 +/- 0.05 mrem/rad (10(-3) Sv/Gy) at 1 m lateral to beam center in the patient midplane at 25 MV. The fraction of this result due to thermal neutrons was found to be only about 2%. The photon background dose was 2.98 +/- 0.04 mrad/rad (10(-3) Gy/Gy). The photoneutron dose equivalent per unit primary dose was found to be nearly independent of the collimator size used but increased by 40% when the bremsstrahlung endpoint energy was increased from 20 to 35 MeV with no change in flattening filters.     

Observations on personnel dosimetry for radiotherapy personnel operating high-energy LINACs

A series of measurements were conducted to determine the cause of a sudden increase in personnel radiation exposures. One objective of the measurements was to determine if the increases were related to changing from film dosimeters exchanged monthly to TLD-100 dosimeters exchanged quarterly. While small increases were observed in the dose equivalents of most employees, the dose equivalents of personnel operating medical electron linear accelerators with energies greater than 20 MV doubled coincidentally with the change in the personnel dosimeter program. The measurements indicated a small thermal neutron radiation component around the accelerators operated by these personnel. This component caused the doses measured with the TLD-100 dosimeters to be overstated. Therefore, the increase in these personnel dose equivalents was not due to changes in work habits or radiation environments. Either film or TLD-700 dosimeters would be suitable for personnel monitoring around high-energy linear accelerators. The final choice would depend on economics and personal preference.     

Dose equivalent response of a TLD badge--influence of body backscatter

In personnel monitoring, operational quantities recommended by ICRU Publication No. 39 for photon radiation can be realized by calibrating dosimeters on a phantom and considering body backscatter photons by using established conversion factors. Personnel dosimeters used in this study are based on CaSO4:Dy Teflon thermoluminescence dosimeter discs (TLD) that have a highly photon energy-dependent response. Since body backscattered photons have lower energies than the incidence photons, methods for correcting for energy dependence of both the incident and body backscattered photons have to be developed. By using readouts of two TLD discs (one under a composite metal filter and the other without a metal filter) in an empirical relation valid at all energies, it is possible to correct for the effect of change in response from change in the photon energies. It was found that the new operational quantities recommended by ICRU could be estimated to within +/- 15% by a TLD badge design based on this method. Angular dependence limits for photons in accordance with the new international standards and a high beta dose-equivalent discrimination in the mixed fields of beta and low-energy x rays could also be achieved.     

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.     

Benchmark test of transport calculations of gold and nickel activation with implications for neutron kerma at Hiroshima.

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a 252Cf fission neutron source to validate the use of the code for the energy spectrum analyses of Hiroshima atomic bomb neutrons. Nuclear data libraries used in the Monte Carlo neutron and photon transport code calculation were ENDF/B-III, ENDF/B-IV, LASL-SUB, and ENDL-73. The neutron moderators used were granite (the main component of which is SiO2, with a small fraction of hydrogen), Newlight [polyethylene with 3.7% boron (natural)], ammonium chloride (NH4Cl), and water (H2O). Each moderator was 65 cm thick. The neutron detectors were gold and nickel foils, which were used to detect thermal and epithermal neutrons (4.9 eV) and fast neutrons (> 0.5 MeV), respectively. Measured activity data from neutron-irradiated gold and nickel foils in these moderators decreased to about 1/1,000th or 1/10,000th, which correspond to about 1,500 m ground distance from the hypocenter in Hiroshima. For both gold and nickel detectors, the measured activities and the calculated values agreed within 10%. The slopes of the depth-yield relations in each moderator, except granite, were similar for neutrons detected by the gold and nickel foils. From the results of these studies, the Monte Carlo neutron and photon transport code was verified to be accurate enough for use with the elements hydrogen, carbon, nitrogen, oxygen, silicon, chlorine, and cadmium, and for the incident 252Cf fission spectrum neutrons.     

Determination of percentage depth dose for 6 and 10 MV x-rays using Ge-doped optical fibre and TLD-100.

Percentage depth doses for 6 and 10 MV x-ray beams from a linear accelerator were measured using approximately 1 cm long (approximately 0.3 mg) Ge-doped optical fibre as a thermoluminescence dosimeter for two field sizes, 5 x 5 and 10 x 10 cm2. The results indicate that the Ge-doped optical fibre dosimeter is in good agreement with the results from a PTW 30001 cylindrical ionisation chamber and TLD-100. For 6 MV x-ray beams we observe that the depth of maximum dose d(max) is 1.5 and 2 cm for field sizes of 5 x 5 and 10 x 10 cm2 respectively. For 10 MV d(max) is 2 cm for a field size of 5 x 5 cm2 and 2.5 cm for a 10 x 10 cm2 field.     

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.     

Application of glow curve analysis methods to improve TLD-100 dose reassessment performance.

This paper presents a method of improving the TLD-100 dose reassessment performance. This method consists of applying numerical analysis techniques for evaluating the TLD-100 phototransferred thermoluminescence (PTTL) glow curve. From this analysis, a simple procedure for estimating the ultraviolet background components usually present in phototransferred thermoluminescence (TL) signals has been established. This procedure has been implemented in a computer program which performs the automatic evaluation of the glow curves and extracts the dose information contained in the PTTL curves. The use of this computer-aided evaluational method has enabled the extension of the working range of estimated absorbed dose down to 0.2-0.5 mGy with very adequate operational quality for doses even below the conventionally admitted lower reestimation limit (approximately 2 mGy). Because TL readout is a destructive process, the ability to reestimate doses can be important in any kind of dosimetric activity, such as operational dosimetry programs. The other commonly used dosimeter, film, uses a nondestructive readout and, therefore, presents some advantages over TLD when dose reassessment is necessary. With the reported improvements in the TLD-100 dose reassessment performance, the full range of absorbed doses covered by film dosimetry can now be reliably reassessed using TLD-100 dosimeters.     

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.     

Preliminary studies to develop a personal dosemeter for use by aircraft crew.

This paper describes preliminary work to develop a cosmic-radiation dosemeter for use by military aircraft crew. The dosemeter is based on a combination of CR-39 etched-track detectors and TLD-700 thermoluminescent detectors. It is intended that the CR-39 be used to assess the neutron dose, while the TLD-700 is used to assess the photon and charged particle dose. The sensitivity of CR-39 to the neutron component of cosmic radiation was estimated by irradiating samples of the plastic at the CERN-CEC High Energy Reference Field Facility. This facility produced a radiation field with a neutron spectrum resembling that of the neutron component of cosmic radiation at typical airflight altitudes. The response of the CR-39 was linear over the range of doses studied (0.2-6.0 mSv) and there was no significant fading in the 6-month period after irradiation. The TLD-700 component of the dosemeter was calibrated using 137Cs gamma rays. The response of the TLD-700 was linear over the range of doses studied (0.01-5.0 mSv) with no significant fading in the 6-month period after irradiation. It was concluded that a combination of CR-39 and TLD-700 detectors would provide an effective cosmic-radiation dosemeter for use by military aircraft crew.     

Neutron dosimetry in the containment of a pressurized water reactor using a neutron-sensitive beta/gamma dosimetry system

In this study the Panasonic UD-802 dosimeter was evaluated as a potential neutron dosimeter for use in the containment of a pressurized water reactor by comparing the results from the UD-802 with remmeter readings. The Panasonic UD-802 dosimeter is used routinely as a beta and gamma dosimeter but due to the natural Li and B in the thermoluminescent materials, it is also sensitive to neutrons. Since a dosimeter's response to neutrons is energy-dependent, proper calibration of the UD-802 in the environment for which it is to be used was an important consideration of the study. To calibrate the system, UD-802 dosimeters were mounted on polyethylene phantoms and irradiated to reference doses at selected locations in containment. The reference doses were determined based on remmeter dose-rate measurements and stay times. The thermoluminescent response of the dosimeters and the reference measurements were used to obtain a response ratio at each location. The average response ratio (unit of dosimeter response per millirem) was 3.7 and all response ratios were within +/-30% of this mean value. Specific characteristics of the UD-802 were also investigated, that is, the effects that dosimeter distance from the phantom and a person's movement through containment have on response. The dosimeter distance from the phantom was found to have a minimal effect on response, but the system was found to be dependent upon the angle of the phantom relative to the reactor core, necessitating a correction in the calibration factor. The overall conclusion of this study was that the Panasonic UD-802 dosimeter can be used for neutron dosimetry in containment of a pressurized water reactor.     

Evaluation of equivalent dose from neutrons and activation products from a 15-MV X-ray LINAC

A high-energy photon beam that is more than 10 MV can produce neutron contamination. Neutrons are generated by the [γ,n] reactions with a high-Z target material. The equivalent neutron dose and gamma dose from activation products have been estimated in a LINAC equipped with a 15-MV photon beam. A Monte Carlo simulation code was employed for neutron and photon dosimetry due to mixed beam. The neutron dose was also experimentally measured using the Optically Stimulated Luminescence (OSL) under various conditions to compare with the simulation. The activation products were measured by gamma spectrometer system. The average neutron energy was calculated to be 0.25 MeV. The equivalent neutron dose at the isocenter obtained from OSL measurement and MC calculation was 5.39 and 3.44 mSv/Gy, respectively. A gamma dose rate of 4.14 µSv/h was observed as a result of activations by neutron inside the treatment machine. The gamma spectrum analysis showed ²⁸Al, ²⁴Na, ⁵⁴Mn and ⁶⁰Co. The results confirm that neutrons and gamma rays are generated, and gamma rays remain inside the treatment room after the termination of X-ray irradiation. The source of neutrons is the product of the [γ,n] reactions in the machine head, whereas gamma rays are produced from the [n,γ] reactions (i.e. neutron activation) with materials inside the treatment room. The most activated nuclide is ²⁸Al, which has a half life of 2.245 min. In practice, it is recommended that staff should wait for a few minutes (several ²⁸Al half-lives) before entering the treatment room after the treatment finishes to minimize the dose received.     

Risks posed by neutron contamination in high energy radiotherapy

High energy X-rays penetrate tissue deeply, depositing most of their energy beyond the skin and shallow tissues. X-rays with energies above 8 MeV may interact to produce neutrons, to which the patient is then exposed. The overall number of neutrons produced is relatively low, because X-rays may interact in a variety ways and reactions producing neutrons are generally less likely. The biological damage inflicted by neutron radiation depends on the energy of the neutrons, and neutrons with energy around 1 MeV may be up to 20 times more damaging than X-rays. Treatment planning systems (TPS) do not...

     

Response of the Panasonic UD803AS TLD in workplace neutron fields.

Between 1983 and 2004, the United Kingdom Ministry of Defence (UK MOD) assessed photon and beta doses using the Panasonic UD803AS thermoluminescent dosemeter (TLD). The Panasonic UD803AS dosemeter is also sensitive to neutrons, and the aim of the present study was to ascertain whether dose assessments made using the Panasonic UD803AS TLD in UK MOD radiation fields include a conservative estimate of any neutron exposures. This would allow Panasonic UD803AS results to be taken as a measure of the total photon, beta and neutron dose when reconstructing dose records for the purpose of the 'Compensation Scheme for Radiation Linked Diseases'. The neutron response of the Panasonic UD803AS TLD was calculated as a function of neutron energy and angle of incidence by using the MCNP5 Monte Carlo computer code, and the results used to predict the response of the dosemeter in UK MOD neutron fields. It was found that the UD803AS dosemeter over-responded in the majority of these fields. It was therefore concluded that historical Panasonic UD803AS dose assessments will, for the most part, include a conservative estimate of any neutron exposure.     

Dosimetry of fission neutrons in a 1-W reactor,UTR-KINKI

The energy spectrum of fission neutrons in the biological irradiation field of the Kinki University reactor, UTR-KINKI, has been determined by a multi-foil activation analysis coupled with artificial neural network techniques and a Au-foil activation method. The mean neutron energy was estimated to be 1.26 +/- 0.05 MeV from the experimentally determined spectrum. Based on this energy value and other information, the neutron dose rate was estimated to be 19.7 +/- 1.4 cGy/hr. Since this dose rate agrees with that measured by a pair of ionizing chambers (21.4 cGy/hr), we conclude that the mean neutron energy could be estimated with reasonable accuracy in the irradiation field of UTR-KINKI.     

In flight negative pion assessments using time-of-flight technique

Time-of-flight measurement was applied to obtain the energy spectra of the neutron emitted from tissue-equivalent plastic (TEP) exposed by pi- mesons. The separation of electrons and muons in beams was attempted. Neutrons from TEP exposed by pi- mesons at P pi = 500 MeV/c were separated from gamma-rays by the NE 213 liquid scintillator and the pulse-height analyzer. The double differential cross sections for neutron energy are obtained. The cross sections of the neutron energy distributions show a preponderance of low energy neutrons and a high energy tail extending to approximately 100 MeV.     

Determining organ dose conversion coefficients for external neutron irradiation by using a voxel mouse model

A set of fluence-to-dose conversion coefficients has been calculated for neutrons with energies <20 MeV using a developed voxel mouse model and Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation used 37 monodirectional monoenergetic neutron beams in the energy range 10⁻⁹ MeV to 20 MeV, under five different source irradiation configurations: left lateral, right lateral, dorsal–ventral, ventral–dorsal, and isotropic. Neutron fluence-to-dose conversion coefficients for selected organs of the body were presented in the paper, and the effect of irradiation geometry conditions, neutron energy and the organ location on the organ dose was discussed. The results indicated that neutron dose conversion coefficients clearly show sensitivity to irradiation geometry at neutron energy below 1 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。