Neutron Dosimetry Research Articles

Preparation of Tb³⁺-doped K₂GdF₅ Used to Neutron Dosimetry

The neutron absorption section of gadolinium atom is very high, so that the K2GdF5 crystals doped with Tb3+ ions have been studied for neutron dosimetry by the thermoluminescence (TL) method. The K2GdF5 crystals doped 2%, 3% and 5% concentrations of Tb3+ ions were synthesized by solid state reaction method. The TL properties were studied after the exposure of the samples to neutron source. The experimental results showed that the TL glow-curves of K2GdF5:Tb3+ has the simple, suitable shape for the dosimetry applications. The TL sensitivity of this material is higher than that of commercial TLD-100 and TLD-900 dosimeters. The thermal fading effect of Tb3+-doped K2GdF5 is very low. The glow curve consists of the main peak at temperature of 202oC. The glow curve deconvolution into individual the peaks showed that the curve had a main peak at 202oC, and kinetic order b = 2. The study results indicated that the K2GdF5 doped with Tb3+ is very remarkable material can be applied in neutron dosimetry field.

Novel techniques for high precision refractive index measurements, and application to assessing neutron damage and dose in crystals

In this work we present novel techniques for high precision index of refraction measurements for transparent crystals, and demonstrate a change from neutron irradiation. Radiation damage affects the structure of material, which can be read out nondestructively in transparent crystals. There is some difference in gamma-ray and neutron interactions which may be useful in characterization. Ionization from gamma rays produces color centers in the material, producing distinct spectral absorption, and some small shift in the index of refraction. Neutrons produce atomic recoils and, while the recoils do some ionization, they have a much greater efficiency for lattice displacement than do gamma rays, and these displacements can have a greater effect on the index of refraction. Using CaF2 crystals exposed to neutron radiation, together with a new high precision technique of detecting changes of index of refraction, we establish proof that this type of measurement can be used to monitor neutron exposure. This can provide a basic study of material changes with radiation and, with calibration of material in known neutron fields, this may even find application to neutron dosimetry.

Fast neutrons detection in CR-39 and DAM–ADC nuclear track detectors

Abstract Fast detection of neutrons in CR-39 and DAM–ADC nuclear track detectors were investigated using new etching conditions. The neutron irradiation is performed using a 5 mCi Am–Be source present at the National Institute of Standards (NIS) of Egypt. Using the new etching condition, irradiated CR-39 samples were etched for 4 h and DAM-ADC samples for 80 min. Suitable analyzing software has been used to analyze experimental data.The dependence of neutrons track density on the neutrons fluence is investigated. When etched under optimum conditions, the relationship between track density and fluence is determined which is found to be linear. Detection efficiency has been represented for both SSNTDs and found to be constant with fluence, which reflects the importance of using CR-39 and DAM–ADC detectors in the field of neutron dosimetry. Linear relationship between track density and effective dose is determined.

Neutron dose evaluation of Elekta Linac at two energies (10 & 18 MV) by MCNP code and comparison with experimental measurements

Medical linear accelerators, besides the clinically high energy electron and photon beams, produce other secondary particles such as neutrons which escalate the delivered dose. In this study the neutron dose at 10 and 18MV Elekta linac was obtained by using TLD600 and TLD700 as well as Monte Carlo simulation. For neutron dose assessment in 2020 cm2 field, TLDs were calibrated at first. Gamma calibration was performed with 10 and 18 MV linac and neutron calibration was done with 241Am-Be neutron source. For simulation, MCNPX code was used then calculated neutron dose equivalent was compared with measurement data. Neutron dose equivalent at 18 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 3.3, 4, 5 and 6 cm. Neutron dose at depths of less than 3.3cm was zero and maximized at the depth of 4 cm (44.39 mSvGy-1), whereas calculation resulted in the maximum of 2.32 mSvGy-1 at the same depth. Neutron dose at 10 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 2.5, 3.3, 4 and 5 cm. No photoneutron dose was observed at depths of less than 3.3cm and the maximum was at 4cm equal to 5.44mSvGy-1, however, the calculated data showed the maximum of 0.077mSvGy-1 at the same depth. The comparison between measured photo neutron dose and calculated data along the beam axis in different depths, shows that the measurement data were much more than the calculated data, so it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry in linac central axis due to high photon flux, whereas MCNPX Monte Carlo techniques still remain a valuable tool for photonuclear dose studies.

Neutronic parameters of a low enrichment core in reactor LR-0 for MSR research

The study of the reactivity effects, which are in many cases negligible, is not a simple task. Consequently, the special core configuration with high sensitivity to low introduced reactivity changes was developed at LR-0 reactor. It proved to be suitable for precise definition of any material, in terms of its impact on reactivity, being inserted into a reactor core. The results also imply this kind of core configuration to be applicable for the planed precise measurement of the MSR FLIBE salt temperature reactivity coefficients. This paper also examines the agreement between calculations and experiments for some cases of the low reactivity introduction into the core. Besides that, the similar cores can be used for purposes of neutron dosimetry, therefore the neutron spectrum in the experimental region was determined and compared with the results of MCNP calculations with the standard nuclear data libraries.

Effects of neutron irradiation on optical and chemical properties of CR-39: Potential application in neutron dosimetry

Effects of high-dose neutron irradiation on chemical and optical properties of CR-39 were studied using FTIR (Fourier Transform Infrared) and UV–vis (Ultraviolet–Visible) spectroscopy. The primary goal was to find a correlation between the neutron dose and the corresponding changes in the optical and chemical properties of CR-39 resulted from the neutron irradiation. The neutrons were produced by bombarding a thick Be target with 22-MeV protons. In the FTIR spectra, prominent absorbance peaks were observed at 1735cm−1 (CO stretching), 1230cm−1(C–O–C stretching), and 783cm−1(C–H bending), the intensities of which decreased with increasing neutron dose. The optical absorbance in the visible range increased linearly with the neutron dose. Empirical relations were established to estimate neutron doses from these optical properties. This technique is particularly useful in measuring high doses, where track analysis with an optical microscope is difficult because of track overlapping.

Characterisation of neutron fields: challenges in assessing the directional distribution.

The SCK·CEN has carried out neutron field characterisation campaigns at several nuclear reactors. The main goal of these measurement campaigns was to evaluate the performance of different neutron personal dosemeters. To be able to evaluate the performance of neutron personal dosemeters in terms of Hp(10), knowledge of the directional distribution is indispensable. This distribution was estimated by placing several personal dosemeters on all six sides of a slab phantom. The interpretation and conversion of this information into a reliable value for Hp(10) requires great care. The data were analysed using three methods. In the first approach, a linear interpolation was performed on three perpendicular axes. In the other two approaches, an icosahedron was used to model the angle of incidence of the neutrons and a linear interpolation or a Bayesian analysis was performed. This study describes the limitations and advantages of each of these methods and provides recommendations for their use to estimate the personal dose equivalent Hp(10) for neutron dosimetry.

Establishing local workplace field correction factors for neutron personal dosemeters.

The present personal neutron dosemeters still need local correction factors to be able to provide accuracy comparable with photon dosemeters. Characterisation of the local neutron field is an indispensable part of neutron dosimetry to obtain such correction factors. It is often overlooked that besides characterisation in the neutron energy also the directional distribution of neutrons plays a crucial part in this characterisation. The authors have done such characterisation in the energy and angle for four workplace fields in Paks NPP. For this a relatively simple approximation method was used using the Nprobe for the energy distribution and measurements on the six sides of the slab phantom with personal dosemeters for the directional distribution. This allowed one to estimate a reference neutron Hp(10) rate and to compare it with the response of several neutron personal dosemeters.

Investigation of the neutron spectrum of americium–beryllium sources by Bonner sphere spectrometry

Americium–beryllium neutron sources are certainly the most widely used in neutron dosimetry laboratories, basically due to their long half-life and their energy distribution, which covers the energy domain of interest for many applications in ambient and personal dosimetry. Nevertheless, the spectrum of this source depends on the materials and dimension of the capsule and on the amount and physical–chemical properties of the active material, thus affecting relevant quantities such as the spectrum-averaged fluence-to-dose equivalent conversion coefficient. A EURAMET (European Association of National Metrology Institutes) project (n. 1104) was initiated to experimentally investigate how the neutron spectrum changes for different source sizes and encapsulations with a view to providing improved data for a planned revision of the ISO 8529 Standard Series. The experimental campaign was carried out in the low scatter facility at NPL. Here three different Bonner sphere spectrometers, BSSs, were exposed to the neutron fields produced by three different neutron sources formats: one X3 capsule (1Ci) and two X14 capsules (10Ci and 15Ci). The specific advantage of the BSS is the large sensitivity to low-energy neutrons (E<0.1MeV) which is the component expected to be most affected by the capsule-to-capsule variations and the component which is least well known.This paper summarises the results of the campaign with emphasis on (1) estimating the low-energy component of the Am–Be neutron spectrum, according to the encapsulation type; (2) evaluating the coherence between the Bonner spheres data and the previous studies performed with high-resolution spectrometers but limited in energy to E>0.1MeV; (3) understanding whether the ISO-recommended Am–Be spectrum needs to be amended, and for which source formats.

Detection and dosimetry studies on the response of silicon diodes to an 241Am-Be source

Silicon diode detectors show potential for the development of an active personal dosimeter for neutron and photon radiation. Photons interact with the constituents of the diode detector and produce electrons. Fast neutrons interact with the constituents of the diode detector and converter, producing recoil nuclei and causing (n,α) and (n,p) reactions. These photon- and neutron-induced charged particles contribute to the response of diode detectors. In this work, a silicon pin diode was used as a detector to produce pulses created by photon and neutron. A polyethylene fast neutron converter was used as a recoil proton source in front of the detector. The total registered photon and neutron efficiency and the partial contributions of the efficiency, due to interactions with the diode and converter, were calculated. The results show that the efficiency of the converter-diode is a function of the incident photon and neutron energy. The optimized thicknesses of the converter for neutron detection and neutron dosimetry were found to be 1 mm and 0.1 mm respectively. The neutron records caused by the (n,α) and (n,p) reactions were negligible. The photon records were strongly dependent upon the energy and the depletion layer of the diode. The photons and neutrons efficiency of the diode-based dosimeter was calculated by the MCNPX code, and the results were in good agreement with experimental results for photons and neutrons from an 241Am-Be source.

SU‐E‐T‐329: Tissue‐Equivalent Phantom Materials for Neutron Dosimetry in Proton Therapy

Purpose:To characterize tissue equivalence of phantom materials in terms of secondary neutron production and dose deposition from neutrons produced in radiation therapy phantom materials in the context of proton therapy using Monte Carlo simulations and measurements.Methods:In order to study the influence of material choice on neutron production in therapeutic proton beams, Monte Carlo simulations using the Geant4 and MCNPX‐PoliMi transport codes were performed to generate the neutron fields produced by protons of 155 and 200 MeV. A simple irradiation geometry was used to investigate the effect of different materials. The proton beams were stopped in slab phantoms to study the production of secondary neutrons. The investigated materials were water, Lucite, and tissue‐equivalent phantom materials (CIRS Inc., Norfolk, VA). Neutron energy spectra and absorbed dose by neutrons and their secondary particles were scored. In addition, simulations were performed for reference tissues (ICRP/ICRU) to assess tissue equivalence with respect to neutron generation and transport. In order to benchmark the simulation results, measurements were performed with a system developed at the University of Michigan; organic liquid scintillators were used to detect the neutron emissions from the irradiation of tissue‐equivalent materials. Additionally, the MPPost code was used to calculate the scintillator response from the MCNPX‐PoliMi output.Results:The simulated energy spectra and depth dose curves of the neutrons produced in different phantom materials showed similar shape. The differences of spectra and fluences between all studied materials and reference tissues were well within the achievable precision of neutron dosimetry. The shape of the simulated detector response of the liquid scintillators agreed well with measurements on the proton beamline.Conclusion:Based on Geant4 and MCNPX‐PoliMi simulations, the investigated materials appear to be suitable to study the production of neutrons in proton therapy. MC simulations were verified with neutron measurements in therapeutic proton beams.This work was funded in part by the ANDANTE grant of the European Commission in the 7th Framework Program.

Comparison of Bonner sphere responses calculated by different Monte Carlo codes at energies between 1 MeV and 1 GeV – Potential impact on neutron dosimetry at energies higher than 20 MeV

Bonner Spheres Spectrometry in its high-energy extended version is an established method to quantify neutrons at a wide energy range from several meV up to more than 1 GeV. In order to allow for quantitative measurements, the responses of the various spheres used in a Bonner Sphere Spectrometer (BSS) are usually simulated by Monte Carlo (MC) codes over the neutron energy range of interest. Because above 20 MeV experimental cross section data are scarce, intra-nuclear cascade (INC) and evaporation models are applied in these MC codes. It was suspected that this lack of data above 20 MeV may translate to differences in simulated BSS response functions depending on the MC code and nuclear models used, which in turn may add to the uncertainty involved in Bonner Sphere Spectrometry, in particular for neutron energies above 20 MeV. In order to investigate this issue in a systematic way, EURADOS (European Radiation Dosimetry Group) initiated an exercise where six groups having experience in neutron transport calculations with the MC codes (MCNP, MCNPX, FLUKA, PHITS, MARS, or GEANT4) calculated the responses of a bare 3He proportional counter, a 3He proportional counter embedded in the middle of a 9 inch polyethylene sphere, and a 3He proportional counter centred in a 9 inch polyethylene sphere containing a lead shell, at neutron energies of 1, 10, 100, and 1000 MeV. In general, calculated responses agreed very well for neutron energies below 20 MeV, whatever MC code used. At higher energies, however, certain differences were observed among the different calculations, which may mainly be attributed to the application of different INC models and their parameters. It was found that up to 1 GeV most of the results ranged between calculations previously published that were obtained with MCNP/LAHET using the Bertini INC model and GEANT4 using the Binary and Bertini INC models. These results indicate that use of different MC codes and INC models for the calculation of BSS response functions may result in an uncertainty of unfolded neutron fluences above 20 MeV of about 20%.

Neutron ESR dosimetry through ammonium tartrate with low Gd content.

This paper continues analyses on organic compounds for application in neutron dosimetry performed through electron spin resonance (ESR). Here, the authors present the results obtained by ESR measurements of a blend of ammonium tartrate dosemeters and gadolinium oxide (5 % by weight). The choice of low amount of Gd is due to the need of improving neutron sensitivity while not significantly influencing tissue equivalence. A study of the effect of gadolinium presence on tissue equivalence was carried out. The experiments show that the neutron sensitivity is enhanced by more than an order of magnitude even with this small additive content. Monte Carlo simulations on the increment of energy release due to gadolinium presence were carried, and the results were in good agreement with the experimental data.

Results of nDOSE and HiDOSE Experiments for Dosimetric Evaluation During STS-134 Mission

HiDOSE (Heavy ion DOSimetry Experiment) and nDOSE (neutron DOSimetry Experiment) experiments conducted as a part of BIOKIS (Biokon in Space) payload were designed to measure the dose equivalent due to charged particles and to neutron field, on the entire energy range, during STS-134 mission. Given the complexity of the radiation field in space environment, dose measurements should be considered an asset of any space mission, and for this reason HiDOSE and nDOSE experiments represent an important contribution to the radiation environment assessment during this mission, a short duration flight. The results of these experiments, obtained using Thermo Luminescence Dosimeters (TLDs) to evaluate the charged particles dosimetry and neutron bubbles dosimeters and stack bismuth track dosimeters for neutron dosimetry, indicate that the dose equivalent rate due to space radiation exposure during the STS-134 mission is in accordance with the results obtained from long duration flights.

State of art: Optically stimulated luminescence dosimetry – Frontiers of future research

Since the commercial adoption of the optically stimulated luminescence (OSL) technique in dosimetry, almost 20 years ago, we have seen major advances in the deployment of OSL dosimeters in different areas, including personal, medical, and space dosimetry. The objective of this paper is to provide a critical overlook at the OSL technique from three different points of view: strengths, challenges and opportunities. We discuss factors that made the OSL technique successful: its simplicity, accuracy, wide dynamic range of measured dose, ease for automation, re-read capability, ability to perform imaging, and the availability of diverse instruments and materials. We look into problems that were overcome and others that remain in several areas of new applications into which OSL has expanded in the past 10 years, such as medical, space, neutron and accident dosimetry. Finally, we discuss unexplored possibilities, new driving forces, and open questions. We hope the broad overview presented here will encourage more discussion and stimulate the research that will advance our fundamental understanding of the OSL process.

Secondary neutron doses received by paediatric patients during intracranial proton therapy treatments

This paper’s goal is to assess secondary neutron doses received by paediatric patients treated for intracranial tumours using a 178 MeV proton beam. The MCNPX Monte Carlo model of the proton therapy facility, previously validated through experimental measurements for both proton and neutron dosimetry, was used. First, absorbed dose was calculated for organs located outside the clinical target volume using a series of hybrid computational phantoms for different ages and considering a realistic treatment plan. In general, secondary neutron dose was found to decrease as the distance to the treatment field increases and as the patient age increases. In addition, secondary neutron doses were studied as a function of the beam incidence. Next, neutron equivalent dose was assessed using organ-specific energy-dependent radiation weighting factors determined from Monte Carlo simulations of neutron spectra at each organ. The equivalent dose was found to reach a maximum value of ∼155 mSv at the level of the breasts for a delivery of 49 proton Gy to an intracranial tumour of a one-year-old female patient. Finally, a thorough comparison of the calculation results with published data demonstrated the dependence of neutron dose on the treatment configuration and proved the need for facility-specific and treatment-dependent neutron dose calculations.

Albedo neutron dosimetry in Germany: regulations and performance

Personal neutron dosimetry has been performed in Germany using albedo dosemeters for >20 y. This paper describes the main principles, the national standards, regulations and recommendations, the quality management and the overall performance, giving some examples.

Use of a spherical albedo system for correcting the readings of albedo dosimeters in JINR phasotron neutron radiation fields

Results of calibrating a spherical albedo system in the radiation fields of a Pu-Be radionuclide neutron source are presented. It is shown that it can be used for correcting the readings of the DVGN-01 albedo dosimeter. The results of measurements with the system in JINR phasotron neutron fields for the purpose of correcting the DVGN-01 readings in these fields are given. The values of the correction factors for DVGN-01 albedo dosimeters when used in personnel neutron dosimetry (PD) on the JINR phasotron are determined.

Benchmark experiments for validation of reaction rates determination in reactor dosimetry

Abstract The precision of Monte Carlo calculations of quantities of neutron dosimetry strongly depends on precision of reaction rates prediction. Research reactor represents a very useful tool for validation of the ability of a code to calculate such quantities as it can provide environments with various types of neutron energy spectra. Especially, a zero power research reactor with well-defined core geometry and neutronic properties enables precise comparison between experimental and calculated data. Thus, at the VR-1 zero power research reactor, a set of benchmark experiments were proposed and carried out to verify the MCNP Monte Carlo code ability to predict correctly the reaction rates. For that purpose two frequently used reactions were chosen: He-3(n,p)H-3 and Au-197(n,γ)Au-198. The benchmark consists of response measurement of small He-3 gas filled detector in various positions of reactor core and of activated gold wires placed inside the core or to its vicinity. The reaction rates were calculated in MCNP5 code utilizing a detailed model of VR-1 reactor which was validated for neutronic calculations at the reactor. The paper describes in detail the experimental set-up of the benchmark, the MCNP model of the VR-1 reactor and provides a comparison between experimental and calculated data.

Tomsk Polytechnic University cyclotron as a source for neutron based cancer treatment

In this paper we present our cyclotron based neutron source with average energy 6.3 MeV generated during the 13.6 MeV deuterons interactions with beryllium target, neutron field dosimetry, and dosimetry of attendant gamma fields. We also present application of our neutron source for cancer treatment.