Photon Dose Rate Research Articles

Measurement and Analysis of Dose Rate Distribution in the Laminated Shielding Assembly Irradiated by High Energy Photons

Benchmark experiments on dose rate of neutrons and photons in the laminated shielding assembly have been performed at the 45 MeV electron linac facility of Hokkaido University. The accelerating energies of electron were 18, 25 and 34 MeV The shielding materials were iron and concrete. Photons were produced in the Cu target set in the direction of 0 degree to the beam line. The detectors used were CR-39 and glass-dosimeter for neutrons and photons, respectively. Analyses have also been performed by using the Monte-Carlo calculation code MCNP-5. The nuclear data libraries used for the analyses were JENDL-3.3, MCPLIB04, and EL3 for neutrons, photons, and electrons, respectively. Three different libraries reproduced by using NJOY process code from LA150, JENDL/PD-2004, and KAERI, were applied for the calculation of photo-nuclear reaction. Comparisons between experimental results and calculated ones were performed. For neutrons very good agreements were obtained at the front surface of the shielding assembly in the case of all energies, however calculational to experimental (C/E) values increased as the depth of the shielding assembly increased. On the other hand, underestimations of the calculated results for photons were observed except on the front surface. It is concluded that further investigation of production of photo-neutrons and secondary gamma-rays in the iron and concrete laminated shield would be necessary.

Measurement and Analysis of Angular Dose Rate Distribution from the Targets Irradiated by A Few Tens of MeV Electrons

In recent years, the acceleration energy of electron linac for medical use increases up to higher than 10 MeV. The angular distributions of photo-neutrons and bremsstrahlung photons emitted from the linear accelerator (linac) have become important issues for the evaluation of exposure dose to whole body of patient and the maze design in linac room. Therefore, benchmark experiments on angular dose rate of neutrons and photons from the metal targets have been performed at the 45 MeV linac facility of Hokkaido University. The electron energies were 18, 25 and 34 MeV. The metal target materials were aluminum, iron, copper, tungsten and lead. The detectors used were CR-39 for neutron and glass-dosimeter for photons. Analyses have also been performed by using the Monte Carlo calculation code MCNP5, for neutron, photon and electron. Three different libraries of LA150, JENDL/PD-2004 and KAERI were applied for photonuclear reaction. Comparisons between experimental results and calculated ones were performed. For neutrons and photons, although the trend of experimental data for angular distribution was well traced by calculation, the underestimation of calculation was conspicuous. It is concluded that further investigation of production of neutrons, photons and electrons from the targets and close components should be necessary.

Radiation exposure in the environment of patients after application of radiopharmaceuticals

Summary Aim: After application of radiopharmaceuticals the patient becomes a radioactive source which leads to radiation exposure in the proximity. The photon dose rates after administration of different radiopharmaceuticals used in diagnostic nuclear medicine were measured at several distances and different time intervals. These data are of importance for estimating the exposure of technologists and members of the public. Patients, method: In this study dose rates were measured for 67 patients after application of the following radiopharmaceuticals: 99mTc-HDP as well as 99mTcpertechnetate, 18F-fluorodeoxyglucose, 111In-Octreotid and Zevalin® and 123I-mIBG in addition to 123I-NaI. The dose rates were measured immediately following application at six different distances to the patient. After two hours the measurements were repeated and – whenever possible – after 24 hours and seven days. Results: Immediately following application the highest dose rates were below 1 mSv / h: with a maximum at 780 μSv/h for 18F (370 MBq), 250 μSv/h for 99mTc (700 MBq), 150 μSv/h for 111In (185 MBq) and 132 μSv/ h for 123I (370 MBq). At a distance of 0.5 m the values decrease significantly by an order of magnitude. Two hours after application the values are diminished to 1/3 (99mTc, 18F), to nearly ½ (123I) but remain in the same order of magnitude for the longer-lived 111In radiopharmaceuticals. Conclusion: For greater distances the doses remain below the limits outlined in the national legislation.

Photon dosimetry using Red 4034 Harwell dosemeters in a pool-type research reactor

The performance of Red 4034 Perspex dosemeters from Harwell Technologies in routine photon dosimetry in mixed fields was evaluated. The dosemeters were irradiated in core grid positions of the Portuguese Research Reactor under a photon dose rate of the order of 10 4 Gy / h , a thermal neutron flux in the range of 1.8 × 10 10 to 1.1 × 10 11 n / cm 2 / s and temperatures below 40 ∘ C . The dosemeters visibly darken upon irradiation and this effect, accurately measurable by spectrophotometry, is related to absorbed dose. Thermoluminescence dosemeters and a calibrated ionization chamber were used as comparison. The results show that Red 4034 Perspex is a suitable dosemeter for mixed-field dosimetry up to 50 kGy .

Analysis of a shield design for a DT neutron generator test facility

Independent numerical simulations have been performed using the MCNP5 and SCALE5 radiation transport codes to evaluate the effectiveness of a concrete facility designed to shield personnel from neutron radiation emitted from DT neutron generators. The analysis considered radiation source terms of 14.1 MeV monoenergetic neutrons located at three discrete locations within the two test vaults in the facility, calculating neutron and photon dose rates at 44 locations around the facility using both codes. In addition, dose rate contours were established throughout the facility using the MCNP5 mesh tally feature. Neutron dose rates calculated outside of the facility are predicted to be below 0.01 mrem/h at all locations when all neutron generator source terms are operating within the facility. Similarly, the neutron dose rate in one empty test vault when the adjacent test vault is being utilized is also less then 0.01 mrem/h. For most calculation locations outside the facility the photon dose rates were less then the neutron dose rates by a factor of 10 or more.

Photon and fast neutron dosimetry using aluminium oxide thermoluminescence dosemeters in a pool-type research reactor

Al 2O 3:Mg,Y thermoluminescence (TL) dosemeters were used to measure photon and fast neutron doses in the mixed radiation field of the Portuguese Research Reactor. The dosemeters were irradiated in core positions under a photon dose rate of the order of 10 4 Gy/h and a fast neutron flux in the range of 10 9–10 11 n/cm 2/s. In order to evaluate the ability of the TL dosemeters for mixed field dosimetry at the research reactor, the measurements were compared with results obtained via conventional methods. The agreement between the different methods is better than 13% for the determination of photon doses and within 5% for the determination of neutron fluxes in mixed fields.

Consequences of removing the flattening filter from linear accelerators in generating high dose rate photon beams for clinical applications: A Monte Carlo study verified by measurement

An Elekta SL-25 medical linear accelerator (Elekta Oncology Systems, Crawley, UK) has been modelled using Monte Carlo simulations with the photon flattening filter removed. It is hypothesized that intensity modulated radiation therapy (IMRT) treatments may be carried out after the removal of this component despite it’s criticality to standard treatments. Measurements using a scanning water phantom were also performed after the flattening filter had been removed. Both simulated and measured beam profiles showed that dose on the central axis increased, with the Monte Carlo simulations showing an increase by a factor of 2.35 for 6MV and 4.18 for 10MV beams. A further consequence of removing the flattening filter was the softening of the photon energy spectrum leading to a steeper reduction in dose at depths greater than the depth of maximum dose. A comparison of the points at the field edge showed that dose was reduced at these points by as much as 5.8% for larger fields. In conclusion, the greater photon fluence is expected to result in shorter treatment times, while the reduction in dose outside of the treatment field is strongly suggestive of more accurate dose delivery to the target.

Energy and Angular Responses of the Criticality Accident Detector using a Plastic Scintillator

The Japan Atomic Energy Agency (JAEA), Nuclear Fuel Cycle Engineering Laboratories, operates a spent fuel reprocessing plant and MOX (Plutonium-Uranium Mixed Oxide) fuel fabrication plants. Criticality accident detectors have been installed in these facilities. The detector, the Toshiba RD120, is composed of a plastic scintillator coupled to a photomultiplier tube, and an operational amplifier. The alarm triggering point is set to 1.0–3.6 mGy.h−1 in photon dose rate to detect the minimum accident of concern. However, a plastic scintillator is principally sensitive not only to primary photons but also to neutrons by secondary photons and heavy charged particles produced in the detector itself. The authors calculated energy and angular responses of the RD120 criticality accident detector to photons and neutrons using Monte Carlo computer codes. The response to primary photons was evaluated with the MCNP-4B and EGS4 calculations, and photon and X-ray irradiation experiments. The response to neutrons that produce secondary photons and heavy charged particles from neutron interactions was computed using the MCNP-4B and SCINFUL, respectively. As a result, reliable response functions were obtained. These results will be a great help in reassessing the coverage area and in determining the appropriate triggering dose rate level in criticality accidents.

Characterisation of PRESAGE™: A new 3-D radiochromic solid polymer dosemeter for ionising radiation

For the past 50 years there has been interest in developing 3-D dosemeters for ionising radiation. Particular emphasis has been put on those dosemeters that change their optical properties in proportion to the absorbed dose. Many of the dosemeters that have been evaluated have had limitations such as lack of transparency, diffusion of the image of the dose distribution or poor stability of baseline optical density. Many of these performance limitations have been overcome by the development of PRESAGE, an optically clear polyurethane-based radiochromic 3-D dosemeter. The solid PRESAGE dosemeter is formulated with a free radical initiator and a leuco dye and it does not require a container to maintain its shape. The polyurethane matrix is tissue equivalent and prevents the diffusion of the dose distribution image. There is a linear dose-response, which is independent of both photon energy and dose rate. Simple precautions such as preventing long-term exposure to additional ionising radiation including ultraviolet and controlling storage temperatures prevent the bleaching of the radiochromic response field within the irradiated dosemeter.

SU‐FF‐T‐380: Radiological Dependence of Electronic Brachytherapy Simulation On Input Parameters

Purpose: In comparison to 125I or 192Ir, characterization of dose rate distributions from electronic brachytherapy is subject to the additional challenge of unforeseen photon energy spectra. Towards simulating photon energy spectra and resultant dose rate distribution, Monte Carlo investigators first generate electrons which bombard the x‐ray tube anode and subsequently create photons via bremsstrahlung. Modeling techniques for this endeavor are largely unexplored. Therefore, sensitivities of spectra and dose rate distributions were assessed through varying modeling parameters for the Xoft Axxent x‐ray source. Materials & Methods: MCNP5 was used to simulate photon spectra and dose rate distributions, with comparisons to experimental measurements (PTW model 34013 chamber in liquid water) for 1<r⩽7 cm and 0°⩽θ⩽150° with simulations covering 0.3 ⩽r⩽ 15 cm and all available angles. The following source modeling parameters were evaluated for impact on in‐water spectra and dose: electron beam radius (R), electron beam annularity (R′) like a doughnut, and anode film thickness (t). Since simulations of electron:photon transport are inefficient in comparison to Monte Carlo modeling of radionuclides, MCNP variance reduction techniques such as cell importances (IMP), electron cutoff energies (PHYS:E), high‐energy biasing of bremsstrahlung spectrum (BBREM), and bremsstrahlung photon multiplicity (BNUM) were assessed. Results: Due to the complex anode shape, F(r,θ) was highly‐dependent on R, varying a factor of 2 when changing R from 0 to 0.084 cm. This effect was more pronounced when varying R′ due to less radial volume averaging. Through comparison with experimental measurements, the optimal electron beam shape had the largest spot size which could fit within the anode and no annularity; it was a uniform pencil beam. Altering MCNP variance reduction techniques did not significantly alter results, but greatly hastened simulation efficiency. Conflict of Interest: Research was sponsored in part by Xoft, Inc.

Superheated drop neutron detectors used at the ESRF

The ESRF operates a 6 GeV electron storage ring to produce X-rays. The experimental hall is classified as a free access area in terms of radiation protection. Interlocked radiation monitors guarantee the corresponding dose constraint. Measurements have shown that the radiation field outside the storage ring is dominated by neutrons. Apfel REMbrandt superheated drop detectors have been chosen. In total, 64 monitors are installed around the storage ring, interlocked to the accelerator personnel safety system. Since the radiation fields are highly pulsed, the response of the REMbrandt monitors to pulsed radiation was measured and compared with theoretical predictions. Dose recordings from the bubble detectors are shown, illustrating the good correlation between beam losses and dose rates, as well as showing the correspondence between neutron and photon dose values. Finally, around the beamline hutches, REMbrandt neutron monitors have been used to evaluate the ratio between neutron and photon dose rates.

Dose rate assessment in complex geometries

A code developed for calculating dose rate maps in scenarios with complex geometries and gamma radiation sources is presented in this work. It is based on the point kernel method. The code computes photon flux and dose rate in preselected positions due to volumetric radioactive sources in the presence of other nonemitting materials acting as shielding. Buildup factors are considered to account for radiation scattering effects, using the Geometric-Progression formula in the fitting function. The major advantage of this code is that there are no constraints for the generation of geometries. Any type and number of volumes can be loaded to generate the scenario by using standard three-dimensional computer-aided design tools. The geometry is imported as a DXF format file. In a second step, material and radiometric attributes are imported for each volume. Volumetric radioactive sources are represented by a set of single emitting point sources distributed within its geometry. The code output has been verified in three different ways: by comparing with results from other point-kernel packages, by comparing with estimates obtained with a Monte Carlo code, and by contrasting with direct experimental measurements. Results of these comparisons are presented and discussed. Finally, intrinsic limitations of the code are discussed

ISFSI site boundary radiation dose rate analyses

Across the globe nuclear utilities are in the process of designing and analysing Independent Spent Fuel Storage Installations (ISFSI) for the purpose of above ground spent-fuel storage primarily to mitigate the filling of spent-fuel pools. Using a conjoining of discrete ordinates transport theory (DORT) and Monte Carlo (MCNP) techniques, an ISFSI was analysed to determine neutron and photon dose rates for a generic overpack, and ISFSI pad configuration and design at distances ranging from 1 to -1700 m from the ISFSI array. The calculated dose rates are used to address the requirements of 10CFR72.104, which provides limits to be enforced for the protection of the public by the NRC in regard to ISFSI facilities. For this overpack, dose rates decrease by three orders of magnitude through the first 200 m moving away from the ISFSI. In addition, the contributions from different source terms changes over distance. It can be observed that although side photons provide the majority of dose rate in this calculation, scattered photons and side neutrons take on more importance as the distance from the ISFSI is increased.

Measurement of photon energy and dose rate

According to ICRP 60, the radiation weighting factor for photon radiation is set equal to one, irrespective of photon energy and dose rate. However, it is well known that the relative biological effectiveness is not constant for all photon energies and also depends on the dose rate. If these dependencies are to be considered, both the photon energy and the dose rate should be measured. This paper describes the extent to which these quantities can be measured for photons with the available instruments. For individual monitoring the existing dosemeters are mainly used to measure the dose, and so values or estimates of both the dose rate and the photon energy are only available for specially designed instruments as additional information. Some passive dosemeters as well as a few direct reading electronic dosemeters can provide information on photon energy. In addition, some of the direct reading electronic dosemeters have the capability to provide dose rate information. For workplace monitoring the dose rate is the main measured quantity, and so dose rate values are always available. Moreover, there are some special instruments available in the market that are designed to measure photon energy.

Dosimetry of in situ activated dysprosium microspheres

This paper presents the results of a study aimed at investigating the dosimetry of stable dysprosium microspheres activated, in situ, by a linac generated photon beam. In phantom measurements of the neutron flux within an 18 MV photon beam were performed using CR-39 detectors and gold activation. The results were used in conjunction with a Monte Carlo computer simulation to investigate the dose distribution resulting from the activation of dysprosium (Dy) microspheres using an 18 MV photon beam. Different depths, lesion volumes and volume concentrations of microspheres are investigated. The linac lower collimator jaws are assumed completely closed to shield the tumour volume from the photon dose. Using a single AP field with 0 × 0 cm2 field size (closed jaws), a photon dose rate of 600 MU min−1 and 80 cm SSD for 10 min, an average dose exceeding 1 Gy can be delivered to spherical lesions of 0.5 cm and higher diameter. The variation of the average dose with the size of the lesion reaches saturation for tumour volumes exceeding 1 cm in diameter. This report shows that the photon beam of a high-energy linac can be used to activate Dy microspheres in situ and, as a result, deliver a significant dose of beta radiation. Non-radioactive Dy microspheres do not have the toxicity and imaging problems associated with commercially available yttrium-90 based products.

Using MCNP Code for Neutron and Photon Skyshine Analysis

The MCNP Monte-Carlo code was used for the investigation of the sensitivity of neutron and neutron-induced secondary photon dose rate, total and thermal neutron fluxes and space-energy distributions to energy and angular distribution of radiation source, to thickness and composition of the ground, air density (including it changing with height), humidities of air and ground, thermalization effects, detector’s dimension and its disposal above the ground level. The calculations were performed with the assumption that the source or released radiation into the atmosphere can be treated as a point source and the source containment structure has a negligible perturbation on the skyshine radiation field.

Active Neutron Personal Dosemeters - A Review of Current Status

A review is presented of the technical feasibility and/or the availability of active neutron personal dosemeters, and the potential for development of such a dosemeter. The review consists of four parts. The first part is an introduction which considers the operational needs for an active neutron personal dosemeter. There does appear to be a need for an electronic neutron personal dosemeter, although it is not established that the demand for such a device is great. The relative magnitude of photon and neutron doses and dose rates in the nuclear industry has changed owing to the increased shielding of the photon component. Changes in fuel burnup and new fuel types have increased neutron exposure. Work on plant refurbishment of decommissioning will also increase neutron dose rates, and for these workplaces there may be sudden dose rate fluctuations in space or time. However, in the UK in 1996 for example, of 12,000 classified workers monitored for neutrons, there were only 100 recorded neutron doses of greater than 1 mSv. In the second part, the physics of personal neutron dosemeters is outlined and summaries are given of detection techniques. It is clear that the development of a practical electronic neutron dosemeter is difficult. Effort has been devoted by many research and development groups over several decades without complete success. The basic difficulty is the requirement to measure very small depositions of energy from neutron radiation which have to be recognised in the presence of photon and electron radiation. This basic difficulty is compounded by the present need to determine lower doses than hitherto. Brief summaries are given of techniques which are being applied to the development of an active neutron personal dosemeter, or are thought to be possible options for the future. The techniques considered are: low pressure (tissue-equivalent) proportional counters, thin semiconductors with event size analysis; two diode devices; multi-diode 'spectrometers'; detector arrays such as charge coupled detector devices; 6Li or 10B sandwich spectrometers; radfets (field effect transistors); direct ion storage devices; memory devices; organic semiconductors; light pipes; GM tubes; organic films; advanced gas detectors; advanced silicon devices; and bubble detectors. Published papers up to mid-1998 have been considered. The third part summarises the performance and metrological requirements which have been proposed for neutron dosemeters by national and international standards and/or regulatory authorities. The fourth part of the review draws conclusions and suggests future action. Of devices presently being studied, it is considered that the more likely candidates for a successful outcome are, in the short term, direct ion storage or bubble detectors, and in the longer term, charge coupled detectors, memory devices or optical fibres.

Retrospective Assessment of Environmental Dose Rates Using Optically Stimulated Luminescence from Al2O3:C and Quartz

As part of our continuing work on assessing the usefulness of natural and artificial dosemeters in retrospective accident dosimetry, the natural photon dose rates within bricks have been measured using the optically stimulated luminescence (OSL) from Al 2 O 3 :C single crystal dosemeters. These were placed in bricks in house walls for two weeks to integrate the environmental dose rates after which their OSL signals were measured. These results were compared with (1) OSL measurements of quartz samples extracted from the bricks and (2) dose rates determined from laboratory measurements of the natural radionuclide concentrations. It is concluded that the excellent OSL characteristics of Al 2 O 3 :C are attractive for use in short-term evaluations of the environmental dose rates. Depth dose profiles were also measured in bricks, after irradiation in the laboratory with 137 Cs gamma rays, using (1) Al 2 O 3 :C single crystals distributed in holes across a brick section and (2) quartz (extracted from the matrix of the brick) taken across the same section. These were compared with typical depth dose profiles found in bricks collected in the Chernobyl accident area and demonstrate retrospectively that the average energy of the initial accident radiation was lower than that of 137 Cs gamma rays.

Neutron and Photon Spectra and Dose Rates Around a Shielding Cask Placed in a Salt Mine to Simulate a Nuclear Waste Package

The dose equivalent rate around a POLLUX storage container with radioactive waste is increased by neutrons and photons backscattered by the host rock when the waste package is transported to its final position in a repository. This was simulated by a smaller shielding cask with a 252Cf line source. Neutron spectral fluence and dose equivalent rates were measured with Bonner spheres above ground and underground and compared with MCNP calculations. There may be the same quantity of backscattered neutrons as of the unscattered component. The photon spectra were determined with an NE213 scintillation spectrometer above ground. Photon dose equivalent rates are derived and compared with calculated ones and dosemeter readings.

Al 2O 3:C as a sensitive OSL dosemeter for rapid assessment of environmental photon dose rates

The use of Al 2O 3:C single crystals as optically stimulated luminescnce (OSL) dosemeters for rapid assessment of the environmental photon dose rate is proposed. It is shown that Al 2O 3:C possesses higher OSL sensitivity than TL sensitivity. In TL measurements thermal quenching is a major problem that crucially depends on the heating rate used and, therefore, the all-optical nature of the OSL procedure is an obvious advantage as it obviates the necessity to heat the material thereby avoiding destruction of the luminescence signal. It is demonstrated that the exceptional high OSL sensitivity and the energy response (equal to that of quartz) make Al 2O 3:C ideal for measuring the environmental dose rates in connection with luminescence dating and retrospective dosimetry using natural materials and ceramics.