Neutron Dose Rates Research Articles

Attenuation of ambient dose equivalent from neutrons by thick concrete, cast iron and composite shields for high energy proton, 3He, 48Ca and 238U ions on Cu targets for shielding design

Data on neutron dose attenuation by thick concrete, cast iron, and cast iron plus concrete composite shields for heavy ions and protons having high energies (200–1000 MeV/u) are necessary for shielding designs of high-powered heavy ion accelerator facilities. Neutron production source terms, shield material attenuation lengths, and neutron dose rate reduction effectiveness of the bulk shielding in the angular range from 0° to 125° were determined by the Particle and Heavy Ion Transport Code (PHITS) for beams of 300 and 550 MeV/u 48Ca ions, 200 and 400 MeV/u 238U ions, 800 MeV/u 3He and 1 GeV protons. Calculated results of interaction lengths of concrete and cast iron were also compared with similar work performed by Agosteo et al., and to experimental and other calculated data on interaction lengths. The agreement can be regarded as acceptable.

Analytic estimates of secondary neutron dose in proton therapy

Proton beam losses in various components of a treatment nozzle generate secondary neutrons, which bring unwanted out of field dose during treatments. The purpose of this study was to develop an analytic method for estimating neutron dose to a distant organ at risk during proton therapy. Based on radiation shielding calculation methods proposed by Sullivan, we developed an analytical model for converting the proton beam losses in the nozzle components and in the treatment volume into the secondary neutron dose at a point of interest. Using the MCNPx Monte Carlo code, we benchmarked the neutron dose rates generated by the proton beam stopped at various media. The Monte Carlo calculations confirmed the validity of the analytical model for simple beam stop geometry. The analytical model was then applied to neutron dose equivalent measurements performed on double scattering and uniform scanning nozzles at the Midwest Proton Radiotherapy Institute (MPRI). Good agreement was obtained between the model predictions and the data measured at MPRI. This work provides a method for estimating analytically the neutron dose equivalent to a distant organ at risk. This method can be used as a tool for optimizing dose delivery techniques in proton therapy.

The dose comparison between the THOR and HFR epithermal neutron beams

Abstract This study is a part of the beam comparison campaign, inter-center dose comparison, between boron neutron capture therapy facilities at the Tsing Hua Open-pool Reactor and the High Flux Reactor. The clinical information exchange can improve the dosimetry uncertainty for medical physics in a mixed field. The method of paired Mg(Ar) and TE(TE) ionization chambers was used to determine the gamma-ray and neutron dose rates. Furthermore, activation foils, including gold, copper, and manganese, were employed to estimate the thermal and epithermal neutron fluxes. Measurements were performed free in air and also in a PMMA phantom. All the chambers were calibrated using a 60Co primary standard source at the Institute of Nuclear Energy Research, Taiwan. Spectrum dependent neutron sensitivity of TE(TE) chamber is one of the important parameters to evaluate dose components. The requested neutron spectra were calculated by the Monte Carlo code MCNP. The measured thermal neutron fluxes, gamma-ray and neutron dose rates of the THOR beam in the phantom were 2.6, 2.2, and 2.1 times of the HFR beam at 2.5-cm depth, respectively. The higher thermal neutron flux and neutron and gamma-ray dose rates are due to the higher epithermal neutron beam intensity of the THOR.

High sensitivity MOSFET-based neutron dosimetry

A new dosemeter based on a metal-oxide-semiconductor field effect transistor sensitive to both neutrons and gamma radiation was manufactured at LAAS-CNRS Laboratory, Toulouse, France. In order to be used for neutron dosimetry, a thin film of lithium fluoride was deposited on the surface of the gate of the device. The characteristics of the dosemeter, such as the dependence of its response to neutron dose and dose rate, were investigated. The studied dosemeter was very sensitive to gamma rays compared to other dosemeters proposed in the literature. Its response in thermal neutrons was found to be much higher than in fast neutrons and gamma rays.

Monte-Carlo Determination of Dose Rates In Spherical Pressurized Water Reactor Shield

The nuclear reactor emits various kinds of nuclear radiations during operation. However, the main contributions to the radiation field in the reactor are given by neutrons and gamma rays. These radiation components are the principal concern of reactor shielding. In our study the neutron and gamma radiation dose rates at different depths in concrete bio-shield of the PWR were calculated based on spherical model for M-C simulation with using MCNP4C2. The simulation results were compared with the results obtained from calculations based on S$_{8}$P$_{3}$ spherical approximation with using the ANISN code.

Flux and dose rate evaluation of iter system using MCNP5

The International Thermonuclear Experimental Reactor (ITER) will perform Deuterium-Tritium (DT) plasma experiments and the neutrons production rate at 14.1 MeV will achieve the level of 1013 n.cm-2.s-1. In this work, the neutron flux and the dose rate during ITER operation has been calculated using the one-dimensional model of the Monte Carlo code MCNP5 and the FENDL/MC-2.1 nuclear data library. The neutron flux and dose rate associated during normal ITER operation were determined along the radial machine direction. Calculations for two different types of concrete compositions were performed to investigate the impact of the bioshield filling materials on the dose rate estimation. The results show that the dose rate level near to the outer wall of the tokamak hall is close to the allowable limit dose. Taking into account the use of large boron concentration in the biological shield concrete (2.9% weight fraction), it was obtained a dose rate reduction by one order of magnitude. The dose rate is dominated by the secondary gamma ray. The magnitude of the dose rate on the outside hall of bioshield during normal ITER operation can not be considered low in accordance with the result found in the simulation performed in this work, i.e., 1 µSv.h-1.

A neutron track etch detector for electron linear accelerators in radiotherapy

BackgroundElectron linear accelerators in medical radiotherapy have replaced cobalt and caesium sources of radiation. However, medical accelerators with photon energies over 10 MeV generate undesired fast neutron contamination in a therapeutic X-ray photon beam. Photons with energies above 10 MeV can interact with the atomic nucleus of a high-Z material, of which the target and the head of an accelerator consist, and lead to the neutron ejection.Results and conclusions.Our neutron dosimeter, composed of the LR-115 track etch detector and boron foil BN-1 converter, was calibrated on thermal neutrons generated in the nuclear reactor of the Josef Stefan Institute (Slovenia), and applied to dosimetry of undesirable neutrons in photon radiotherapy by the linear accelerator 15 MV Siemens Mevatron. Having considered a high dependence of a cross-section between neutron and boron on neutron energy, and broad neutron spectrum in a photon beam, as well as outside the entrance door to maze of the Mevatron, we developed a method for determining the effective neutron detector response. A neutron dose rate in the photon beam was measured to be 1.96 Sv/h. Outside the Mevatron room the neutron dose rate was 0.62 μSv/h. PACS: 87.52. Ga; 87.53.St; 29.40.Wk.

Diagnostic assessment to estimate and minimize neutron dose rates received by occupationally exposed individuals at cyclotron facilities

Since 2003, radiopharmaceuticals for medical diagnostic purposes have been produced at the Instituto de Engenharia Nuclear, in Brazil, using two cyclotron accelerators - CV-28 and RDS111. As a result of the ever increasing production, a diagnostic assessment to reduce neutron dose rates received by occupationally exposed individuals during irradiation processes has been developed. The purpose of this work is to present this assessment, which is currently being applied to both the Fluorine and Iodine targets of CV-28 and RDS111 cyclotron accelerators.

Decay Gamma Dose Rates in the EVEDA Accelerator: Impact of the Deuteron Loss Uncertainties in Accelerator Maintenance

Before starting with the construction of the International Fusion Materials Irradiation Facility, the objective of the Engineering Validation and Engineering Design Activities (EVEDA) phase will be to demonstrate feasibility of design. For this EVEDA prototype, analysis of the dose rate evolution during the beam-off phase is necessary for radioprotection and maintenance feasibility purposes. The key points for determining the dose rates of the beam-off phase are on one hand the neutron source produced along the accelerator beam line and on the other hand the deuteron losses giving rise to this neutron source.A new methodology to compute the neutron source coming from the deuteron interactions with the intercepting material as well as with the deuterium previously implanted has been developed. This new procedure consists of evaluation of the low-energy deuteron-induced neutron source that is not calculated by most transport codes and assessment of the deuterium concentration evolution in the material, which is generally not taken into account in this type of calculation. The impact of this new approach on the neutron source and dose rate results is very relevant.In addition, different sets of deuteron losses computed during the last 3 yr have been compared and used for neutron activation and dose rate calculations. The effect of the deuteron losses upon neutron source production and residual dose rates in the vicinity of the accelerator components is evaluated, and implications for hands-on maintenance activities are discussed. The impact of the differences in the sets of deuteron losses has been found to be very important for dose rate evaluations. Using the most recent deuteron loss information, we obtain dose rate values more than one order of magnitude lower than those obtained using the former data.

The Recent Improvement and Verification of DARWIN: Development of a New DAQ System and Results of Flight Experiment

To improve radiation safety in high-energy accelerator facilities, the authors have been developing the new radiation dose monitor device DARWIN: Dose monitoring system Applicable to various Radiations with WIde energy raNges. DARWIN is composed of (a) a phoswitch-type scintillation detector, which consists of liquid organic scintillator BC501A coupled with ZnS(Ag) scintillation sheets doped with 6Li, and (b) a data acquisition (DAQ) system for digital analysis of the waveform of the scintillator signals. The DAQ system was recently updated in order to apply DARWIN in monitoring dose rates in radiation fields having time structure, introducing an originally developed module based on a field-programmable gate array. To examine the performance of DARWIN placed in radiation fields composed of varieties of particles over wide energy ranges, the authors mounted DARWIN on a jet aircraft and measured neutron, photon, muon, electron, and positron dose rates at high altitudes. The measured dose rates excellently agreed with the corresponding data calculated by EXPACS over a wide altitude range. This agreement indicates the applicability of DARWIN to dose monitoring in complex radiation fields such as those in high-energy accelerator facilities and aircrafts.

Applicability of convex hull in multiple detector response space for neutron dose measurements

A novel neutron dose measurement method that flexibly responds to variations in the neutron field is being developed by Japan Atomic Energy Agency. This is an implementation of the multi-detector method (first introduced in 1960s) for neutron dose evaluation using a convex hull in the response space defined for multiple detectors. The convex hull provides a range of possible neutron dose corresponding to the incident neutron spectrum. Feasibility of the method was studied using a simulated response of mixed gas proportional counter. Monochromatic neutrons are shown to be fundamentally suitable for mapping the convex. The convex hull can be further reduced taking into consideration a priori information about physically possible incident neutron spectra, for example, theoretically derived moderated neutron spectra originated from a fission neutron source.

Characterization of a boron neutron capture therapy beam line at the University of Missouri Research Reactor

A new research effort aimed at increasing the catalog of boron delivery agents for boron neutron capture therapy (BNCT) has been undertaken by the International Institute of Nano and Molecular Medicine (IINMM) at the University of Missouri. The MU Research Reactor (MURR) and Idaho National Laboratory (INL) have constructed a facility for small animal BNCT experiments. The facility incorporates silicon and bismuth single crystal filters to produce a thermal neutron beam. The thermal beam has a measured thermal flux of 8.8 × 108 n cm−2 s−1 and a gold cadmium ratio of 130. The neutron and photon dose rates were measured using paired ion chambers.

SU‐FF‐T‐61: Determination of Neutron and Gamma Dose Rates in Water Surrounding a New Interstitial Cf‐252 Brachytherapy Source

Purpose: A new generation of high‐activity miniature 252Cf sources for interstitial brachytherapy has recently been encapsulated at the Oak Ridge National Laboratory (ORNL). The purpose of this study is to determine the neutron and gamma dose rates in water surrounding the new 252Cf source. Method & Materials: The neutron and gamma dose components of the new 252Cf source in water were determined by the measurements of two miniature ion chambers (T1 and M1 of Standard Imaging, Inc.) that have similar gamma responses but very different neutron responses. A Lucite‐walled water phantom was built to conduct the experiment. The experimental setup was also modeled by the Monte Carlo transport code MCNP. Results: The results show that the measured neutron absorbed dose rates were approximately 25% lower than that predicted by MCNP for all dose positions in water. The measured gamma absorbed dose rates in water, on the contrary, are higher than that predicted by MCNP. The differences between the measured and MCNP‐predicted gamma dose rates are most pronounced (∼a factor of two) at the distance of 1 cm, and fall to approximately 30% at distances 2 cm and beyond. Conclusions: The measured neutron and gamma dose rates do not agree with the dose rates predicted by MCNP. The discrepancies between the two results suggest that the true 252Cf content of the new source is approximately 25% less than the ORNL estimate and that the new source emits significantly more low‐energy gamma rays than the large 252Cf sources (e.g. the AT source) used in the previous studies. Acknowledgement: This research is supported by Isotron Inc.

SU‐EE‐A1‐03: The Neutron Dose Evaluation and Shielding at the Maze Entrance of the Varian Clinac 2300EX

Purpose: After the installation of a Varian Clinac 2300EX accelerator, neutron and photon doses at the outer maze entrance were measured and compared with several empirical calculations. The measurements were taken before and after borated polyethylene (BPE) boards were installed on the maze wall as neutron absorption material. Method and Materials: The accelerator delivered an 18 MV photon beam at 600 cGy/min dose rate. An NRC NP‐2 type neutron REM‐meter and a Ludlum 14C survey meter with a 44‐6 type of probe were used to measure the neutron dose and photon dose, respectively. The measurements were taken at the center axis of the maze, 0.8 m above the floor, 0.3 m away from the maze door. Results: With the gantry head tilted close to the inner maze entrance with closed jaws, both neutron dose and photon dose reach their maximum. The measurement data were compared with empirical calculation results obtained by Kersey's method, modified Kersey's method and a newly proposed method by R.C. Falcão el al. The estimation from Kersey's method is about 2 – 4 times of the measurement (Ratio ≈ 2.4–3.8). Falcão's method has estimation about 22 times of the measurement (Ratio ≈ 21.9). The modified Kersey's method has the best prediction of the dose (Ratio ≈ 1.0). The McGinley's photon dose equation gives estimation about 80% of the measurement. After applying borated polyethylene boards as lining on maze wall, the neutron dose and the photon dose at maze entrance were decreased by 41% and 58%, respectively. Conclusion: This work indicates that Kersey's method overestimates the neutron dose by about 2 to 4 times for this study. Falcão's method largely overestimates the neutron dose. The modified Kersey's method is recommended to be used in neutron shielding calculation for the 2300EX accelerator.

Design of a laboratory for experiments with a pulsed neutron source

We present the results of a neutron shielding design and optimisation studyperformed to reduce the exposure to radiological doses arising from a 14 MeVpulsed neutron generator (PNG) having a maximum emission strength of2.0 × 108 neutrons s−1. The source was intended to be used in a new irradiation facility for the realisation of anexperiment on acoustical cavitation in liquids. This paper describes in detail how thefacility was designed to reduce both neutron and gamma-ray dose rates to acceptablelevels, taking into account the ALARP principle in following the steps of optimisation. Inparticular, this work compares two different methods of optimisation to assessneutron dose rates: the use of analytical methods and the use of Monte Carlosimulations (MCNPX 2.4). The activation of the surrounding materials duringoperation was estimated using the neutron spectra as input to the FISPACT 3.0code. The limitations of a first-order analytical model to determine the neutronactivation levels are highlighted. The impact that activation has on the choice of thematerials to be used inside the laboratory and on the waiting time before anyone cansafely enter the room after the neutron source is switched off is also discussed.

New empirical formula for neutron dose level at the maze entrance of 15 MV medical accelerator facilities

An easily applicable empirical formula was derived for use in the assessment of the photoneutron dose at the maze entrance of a 15 MV medical accelerator treatment room. The neutron dose equivalent rates around the Varian medical accelerator head calculated with the Monte Carlo code MCNPX were used as the source term in producing the base data. The dose equivalents were validated by measurements with bubble detectors. Irradiation geometry conditions expected to yield higher neutron dose rates in the maze were selected: a 20 x 20 cm2 irradiation field, gantry rotation plane parallel to the maze walls, and the photon beams directed to the opposite wall to the maze entrance. The neutron dose equivalents at the maze entrance were computed for 697 arbitrary single-bend maze configurations by extending the Monte Carlo calculations down to the maze entrance. Then, the empirical formula was derived by a multiple regression fit to the neutron dose equivalents at the maze entrance for all the different maze configurations. The goodness of the empirical formula was evaluated by applying it to seven operating medical accelerators of different makes. When the source terms were fixed, the neutron doses estimated from the authors' formula agreed better with the corresponding MCNPX simulations than the results of the Kersey method. In addition, compared with the Wu-McGinley formula, the authors' formula provided better estimates for the mazes with length longer than 8.5 m. There are, however, discrepancies between the measured dose rates and the estimated values from the authors' formula, particularly for the machines other than a Varian model. Further efforts are needed to characterize the neutron field at the maze entrance to reduce the discrepancies. Furthermore, neutron source terms for the machines other than a Varian model should be simulated or measured and incorporated into the formula for accurate extended application to a variety of models.

Extension of CASCADE.04 to estimate neutron fluence and dose rates and its validation

Capability to compute neutron dose rate is introduced for the first time in the new version of the CASCADE.04 code. Two different methods, ‘track length estimator’ and ‘collision estimator’ are adapted for the estimation of neutron fluence rate needed to calculate the ambient dose rate. For the validation of the methods, neutron dose rates are experimentally measured at different locations of a 5Ci Am-Be source, shielded in Howitzer-type system and these results are compared with those estimated using (i) modified CASCADE.04.d and (ii) MCNP4A codes and it is found that the agreement is good. The paper presents details of modification and results of the comparative study.

Nationwide measurements of cosmic-ray dose rates throughout Japan

Cosmic-ray dose rates on the ground were measured throughout Japan. Neutron dose rates were measured as ambient dose equivalent rates (H*(10)) at 240 points using high-sensitivity neutron REM counters. In addition, cosmic rays directly ionising plus photon components were measured with an ionisation chamber. Time variation due to solar modulation during this study was corrected based on the results of sequential measurements. The effects of altitude, geomagnetic latitude, rainfall and snowfall on the neutron dose rate were inferred from the measured results. The mean value of the neutron dose rates (H*(10)) measured at 47 points of prefectural capitals in Japan was 4.0 nSv h(-1). The value corrected for the energy response of the REM counter was 6.4 nSv h(-1), corresponding to 4.8 nSv h(-1) as an effective dose (ISO). The mean value of the cosmic ray directly ionising plus photon components as an effective dose was 31 nSv h(-1).

The munich fission neutron therapy facility MEDAPP at the research reactor FRM II

At the new research reactor FRM II of the Technical University of Munich (TUM), the facility for Medical Applications (MEDAPP) was installed where fast neutrons are available as a beam for medical use. Thermal neutrons induce fission in a pair of uranium converter plates and generate fast neutrons which are guided to the patient by a beam tube. The maximum opening of the multi leaf collimator (MLC) is 30x20 cm2 WxH. The beam is characterized by neutron-photon mixed beam phantom dosimetry. Specific safety measures are outlined. The neutron and gamma dose rates are 0.52 Gy/min and 0.20 Gy/min, respectively, in 2 cm depth of a water phantom. The half maximum depth of the neutron dose rate in water is 5.4 cm (mean neutron energy 1.9+/-0.1 MeV). Conformity with the European Medical Devices Directive (MDD) 93/42/EEG, was proven so that MEDAPP has a CE mark and since February 2007 also the license for clinical operation. The clinical neutron irradiations of malignant tumors, which were performed at the former research reactor FRM until 2000, can be continued at FRM II under improved conditions. First patients were irradiated in June 2007.

Comparison of Three Pulse Processing Systems for Microdosimetry

Three different pulse processing systems coupled to a tissue-equivalent proportional counter were tested in performance for various neutron and gamma-ray dose rates. The three systems used are a conventional analogue system, a real time system and a digital pulse processing system. The Tandetron accelerator at the McMaster Accelerator Laboratory was used to produce a mixed neutron-gamma field via the 7Li(p,n) reaction. The proton energy ranged from 1.8 to 2.5 MeV with the current set at 100, 200 and 300 muA. The digital system displayed the fastest pulse processing speed. Microdosimetric spectra from the three systems were consistent and both neutron and gamma-ray dose rates showed good agreement in most cases. At neutron dose rates higher than 20 mGy/min the data from the digital system was reliable while systematic deviations were observed for the other two systems.