Neutron-gamma Radiation Fields Research Articles

Tungsten-based polymer composite, a new lead-free material for efficient shielding of coupled neutron-gamma radiation fields: A FLUKA simulation study

Abstract Metal-based polymer composites, a new category of advanced materials, are advantageous for effective protection of radiation field. Recent report of fabrication of tungsten (W)-Poly methyl methacrylate (PMMA) composite microcellular foams with enhanced mechanical strength properties opens up the possibility of its use in radiation attenuation. Objective of this theoretical study is to assess the efficacy of W-based polymer composite, a new lead-free shielding material for attenuating coupled neutron-gamma radiations. Current paper utilizes open-source Monte Carlo code FLUKA to evaluate shielding efficiency of PMMA composites reinforced with varying concentration W particles. Study shows that, adding even 20 vol% of W particles can significantly improve radiation shielding ability of PMMA. Performance of analogous composition Pb-based polymer composite is also examined to demonstrate its inadequacy in radiation protection compared to W-based composite. Study reveals an interesting fact that for any shield dimension, total radiation dose follows an initial descending trend with increase in heavy metal (W/Pb) proportion up to certain optimum value where dose becomes minimum, beyond that dose increases. Optimum heavy metal concentrations are found to be 70 vol% and 30 vol% for W and Pb respectively, with minimum dose for Pb shield being two orders of higher magnitude. Study is further extended to investigate shielding efficiency of conventional double-layer laminates employing W and PMMA in both high-Z/low-Z and low-Z/high-Z configurations as well as optimum concentration of W-PMMA composite and PMMA. It is shown that among all the potential designs, newly introduced composite-based double-layer shield performs best in terms of volumetric dose while single-layer optimized composite shield offers least specific dose.

A new Multi Leaf Collimator for the fission neutron therapy facility at FRM II

As a part of the general upgrade program of the fission neutron therapy facility MEDAPP at the research reactor FRM II, a newly designed, electronically operated Multi Leaf Collimator (MLC) was introduced. The design of the new collimator was optimized in order to increase dose output and to allow patient positioning closer to the collimator. The new collimator allows divergent and convergent beam shaping in order to maximize the effective area of the fission source used for irradiation. In comparison to the old MLC, the radiation dose for the staff is reduced by the motorized operation of the new MLC. The advanced beam shaping abilities of the new collimator are investigated by a comparison with depth dose curves and beam profiles of the old MLC. For the separation of neutron and gamma dose deposition in water, the two chamber dosimetry method recommended for mixed field dosimetry was applied. A re-calibration of the set of ionization chambers used for dosimetry in mixed neutron-gamma radiation fields at FRM II was performed. The calibration was done in a reference neutron field from 3H(p,n)3He-reactions in a low scattering environment. An overall increased dose output as intended by the design of the new MLC was shown. Also, improved dosimetric characteristics of the treatment field due to the optimized beam shaping abilities of the new MLC were demonstrated. In the calibration measurements of the magnesium ionization chamber used in the two chamber method, an unexpected increase of the neutron sensitivity by a factor of 8 was measured. The new MLC will enable more conformal dose delivery to the target volume in patients and therefore help to reduce healthy tissue damage by the application of neutron radiation.

Neutron spectrometry and dosimetry using a multi-shell Single Sphere Neutron Spectrometer with thermo-luminescent and optically stimulated luminescent detectors

A portable neutron spectrometer with concentric multi-shell moderating assembly and TL, OSL detectors has been validated with standard neutron reference fields to generate the neutron spectra and estimate the ambient dose equivalents, H*(10). Further, the system has been used in the mixed neutron–gamma radiation fields at accelerator facility. The neutron fluence and the ambient dose equivalent estimates were verified using the standard reference neutron fields, viz. 241Am-Be, 252Cf, D-T neutron sources. The system was then employed in the accelerator radiation environment to measure the neutron spectral yield from the interactions of 15 and 20 MeV protons with a thick natural Ta target. The results were compared with those obtained from FLUKA simulations. A conventional neutron dose equivalent meter was also used to verify the ambient dose equivalent estimated from the spectrometric measurements. The reproducibility of the neutron spectra and dose values were found to be satisfactory, thereby encouraging the implementation of the portable single sphere neutron spectrometer in complex radiation environment around particle accelerators. Present studies demonstrate the measurements with the in-house built spectrometer for neutron energies up to 20 MeV.

Shielding Capability Research on Composite Base Materials in Hybrid Neutron-Gamma Mixed Radiation Fields

The 16N monitoring system operates in a mixed neutron-gamma radiation field and is subject to high background radiation, thus triggering instability in the 16N monitoring system measurement data. Due to its property of actual physical process simulation, the Monte Carlo method was adopted to establish the model of the 16N monitoring system and design a structure-functionally integrated shield to realize neutron-gamma mixed radiation shielding. First, the optimal shielding layer with a thickness of 4 cm was determined in this working environment, which had a significant shielding effect on the background radiation and improved the measurement of the characteristic energy spectrum and the shielding effect on neutrons was better than gamma shielding with the increase in the shield thickness. Then, functional fillers such as B, Gd, W, and Pb were added to the matrix to compare the shielding rates of three matrix materials of polyethylene, epoxy resin, and 6061 aluminum alloy at 1 MeV neutron and gamma energy. The shielding performance of epoxy resin as the matrix material was better than that of the aluminum alloy and polyethylene, and the shielding rate of boron-containing epoxy resin was 44.8%. The γ-ray mass attenuation coefficients of lead and tungsten in the three matrix materials were simulated to determine the best material for the gamma shielding performance. Finally, the optimal materials for neutron shielding and gamma shielding were combined, and the shielding performance of single-layer shielding and double-layer shielding in mixed radiation field was compared. The optimal shielding material-boron-containing epoxy resin was determined as the shielding layer of the 16N monitoring system to realize the integration of structure and function, which provides a theoretical basis for the selection of shielding materials in a special working environment.

A mm3 Fiber-Coupled Scintillator for In-Core Thermal Neutron Detection in CROCUS

This article presents the development, full-characterization, and in-core testing of a miniature neutron detector for the investigation of highly localized in-core neutronics effects in the zero-power reactor CROCUS, operated at the Ecole Polytechnique federale de Lausanne (EPFL), Switzerland. A ZnS: 6LiF screen, mixing inorganic scintillator and neutron converter, with a surface of 1 mm2 and a thickness of 0.2 mm, was coupled with a silicon photomultiplier (SiPM) via a 10-m optical fiber. Analog readout electronics were adopted in this prototype version to process the SiPM output signal. A first campaign was carried out to assess the detection system capabilities in terms of neutron detection and counting in mixed neutron-gamma radiation fields. The miniature detector was thoroughly characterized with a Pu-Be source installed in the CARROUSEL facility and was subsequently tested inside a control rod guide tube of the CROCUS reactor in different reactor conditions, i.e., at shutdown, startup, and critical. The detector shows a linear response with a reactor power increase up to 6.5 W (i.e., around $10 {^{\mathrm{ 8}}}\,\,\text {cm}^{-2}\cdot \text {s}^{-1} $ total neutron flux) and excellent neutron counting capabilities, when compared to other localized detection systems available in CROCUS, such as the miniature fission chambers and single-crystal vapor deposited (sCVD) diamond detectors. In addition, the exposure of the miniature detector to a 60Co source with an activity of ~250 GBq, combined with the computation of the gamma flux in CROCUS with the Serpent 2 Monte Carlo code, confirmed the gamma insensitivity of the system in a mixed neutron-gamma field.

Design, fabrication and modeling of an AmBe neutron irradiator for TLD screening for neutron dose measurement in mixed radiation fields

TLDs dosimeters are frequently presented as a viable choice for dosimetric studies when dealing with mixed neutron-gamma radiation fields. However, this choice is not without some drawbacks, because not only TLD response is highly dependent on particle type but also on neutron energy spectrum. Therefore, a correct screening and calibration of the dosimeter are required, and a simple shift from gamma screening methodology for mixed field is not suitable. This paper presents the design, fabrication and tests of an irradiator for TLD screening for neutron dose measurement using an AmBe source and polyethylene as moderator material. The design of the irradiator was conducted through Monte Carlo simulations using the MCNP5 code. The experimental validation and tests were performed using Indium activation foils and TLD 600 dosimeters. The manufactured irradiator demonstrated to be suitable for TLD screening under neutron source radiation field, offering very good homogeneity conditions in the radiation field so to guarantee same radiation dose delivered to the TLDs.

Application of Prompt Self-Powered Neutron Detectors to the Lead-Cooled Fast Reactor Demonstrator ALFRED: Validation of the Monte Carlo Model for Selected SPNDs

The advanced lead fast reactor European demonstrator (ALFRED) is a European research initiative into the framework of the Generation IV International Forum facilities. ALFRED is a scaled down reactor compared to the industrial prototype European lead fast reactor proposed in lead-cooled European advanced demonstration reactor. It has a relatively low power (125 MWe) with a compact design to reduce the cost but maintaining its representativeness and it is cooled by pure lead. One of the open issues is linked to the neutron flux in-core monitoring system because of the harshness of the environment the detectors should be installed in, due to high temperatures, and the neutron-gamma radiation field levels. Monte Carlo simulation is a possible way of facing the problem, reproducing into a virtual world the reactor core, the surrounding environment and radiation interactions. In previous works, neutron spectra and gamma doses at possible detectors' locations in ALFRED were retrieved, with consideration on the applicability of each suitable device currently available. Fission chambers (FCs) were found to be exploited at reactor start-up and intermediate power range. Prompt self-powered neutron detectors (SPNDs) seemed to be the best solution to monitor the reactor full power, becoming the main research target: their effective applicability on field has to be demonstrated. SPND applications do not include reactor control purposes usually. Moreover, their irradiation experience involved thermal and epithermal neutron spectra monitoring, mainly. The lack of data when SPNDs sense fast neutron fluxes in terms of prompt-response pushed the authors to deepen the study in such direction. The work herein shows the mathematical approach based on Monte Carlo simulation of SPNDs by the Monte Carlo N-particle eXtended code (MCNPX), so as to study the capability of the code in reproducing real devices' signals while experimented on field. Such a verification turned out to be the preliminary stage for studying new concepts for SPNDs, in terms of sensitive materials and geometries, envisaging the possibility for designing, prototyping, and testing new devices in suitable fast neutron-flux facilities.

Development of a multi-element microdosimetric detector based on a thick gas electron multiplier

A prototype multi-element gaseous microdosimetric detector was developed using the Thick Gas Electron Multiplier (THGEM) technique. The detector aims at measuring neutron and gamma-ray dose rates for weak neutron-gamma radiation fields. The multi-element design was employed to increase the neutron detection efficiency. The prototype THGEM multi-element detector consists of three layers of tissue equivalent plastic hexagons and each layer houses a hexagonal array of seven cylindrical gas cavity elements with equal heights and diameters of 17mm. The final detector structure incorporates 21 gaseous volumes. Owing to the absence of wire electrodes, the THGEM multi-element detector offers flexible and convenient fabrication. The detector responses to neutron and gamma-ray were investigated using the McMaster Tandetron 7Li(p,n) neutron source. The dosimetric performance of the detector is presented in contrast to the response of a commercial tissue equivalent proportional counter. Compared to the standard TEPC response, the detector gave a consistent microdosimetric response with an average discrepancy of 8 % in measured neutron absorbed dose. An improvement of a factor of 3.0 in neutron detection efficiency has been accomplished with only a small degradation in energy resolution. However, its low energy cut off is about 6keV/μm, which is not sufficient to measure the gamma-ray dose. This problem will be addressed by increasing the electron multiplication gain using double THGEM layers.

LiCaAlF6 scintillators in neutron and gamma radiation fields

Intentionally doped LiCaAlF6 (LiCAF) single crystals are prospective scintillators, especially for thermal neutron detection through the 6Li(n,t)4He nuclear reaction. Four different LiCAF scintillator samples were tested in various neutron and gamma fields. Two of the tested samples were LiCAF:Eu and LiCAF:Eu,Na single crystals, and another two samples were made of LiCAF:Eu micro crystals dispersed in transparent rubber, with different rubber dimensions. All LiCAF samples contain lithium enriched to[Formula: see text]Li. A plutonium–beryllium source was used as a neutron source. The neutron spectrum was modified by moderator and filter to get different ratios between thermal, epithermal and fast neutron fluence rates. The MCNP code was used for calculations of the fluence rates for different configurations. Radionuclides [Formula: see text]Cs and [Formula: see text]Co were applied as gamma radiation sources. The light signal from the scintillator was evaluated with a photomultiplier and a multichannel analyzer. The purpose of this work was to study the characteristics of LiCAF scintillators, especially the ability to discriminate signals from neutron and gamma radiation, which is the basic scintillator condition for neutron detection in mixed neutron-gamma radiation fields. Generally, the discrimination can be done by the pulse height and/or the pulse shape of the evaluated signals. Both methods can be used for a LiCAF scintillator. However, only the pulse height discrimination method is discussed in this paper. The possibility of fast neutron detection with LiCAF scintillators was also tested.

A Novel Detector for Measuring Gamma-Ray Fluxes in a Mixed Pulsed Neutron-Gamma Radiation Fields

An ICI (insulator-conductor-insulator) gamma-ray detector designed for measurement of fast rising, high intensity pulsed gamma-ray fluxes has been developed. Like vacuum Compton diode (VCD) and dielectric Compton diode (DCD) devices, the ICI detector operates by utilization of the Compton effect. It has a very fast time response (rise time is less than 1 ns), a large linearity and a wide dynamic range. It is designed to measure intense and rapidly changing gamma-ray fluxes in highly mixed neutron-gamma radiation fields. Its sensitivity to 1.25 MeV gamma rays is about 1.8 x 10-20, for the 50 mm collimation diameter, which is relatively high compared with VCD and DCD devices. The simulated gamma-ray sensitivity agrees well with the experimental value. The gamma-to-neutron sensitivity ratio is about three orders of magnitudes for neutron energies below 14.1 MeV and 1.25 MeV gamma rays. Compared to existing VCD and DCD devices, the ICI detector requires no vacuum and power supply in operation, and also minimally attenuates a collimated gamma-ray beam due to its overall thickness of 3 mm in beam path.

Intercomparison on Measurements of the Quantity Personal Dose Equivalent Hp(d) in Mixed (Neutron-Gamma) Fields--an IAEA Project

Within its occupational radiation protection programme, the International Atomic Energy Agency (IAEA) initiated and funded an international intercomparison exercise of personal dosemeters to determine the quantity personal dose equivalent in mixed neutron-photon radiation. The objectives of the intercomparison are to assess the capabilities of the dosimetry services in measuring the quantity Hp(10) in mixed neutron-gamma fields; to assist IAEA member states in achieving sufficiently accurate dosimetry; and, if necessary, to provide guidelines for improvements (not simply a test of the performance of the existing dosimetry service). The intercomparison is directed to passive dosemeters to determine, in mixed neutron-gamma radiation fields, either these two components separately or the total personal dose equivalent. The intercomparison consists of two phases: Phase I--Type-test intercomparison: irradiation in selected calibration fields and results used to improve dosimetric procedures of participating laboratories, where needed. Phase II--Simulated workplace field intercomparison: irradiation in radiation fields similar to those in workplaces as a final check of performance. The exercise revealed clear deficiencies in the methodology used by several laboratories and necessitated a detailed analysis of the existing discrepancies. This papers summaries the finding and conclusions for radiation fields similar to those found in nuclear industry.

Study of the relative dose-response of BANG-3® polymer gel dosimeters in epithermal neutron irradiation

Polymer gels have been reported as a new, potential tool for dosimetry in mixed neutron–gamma radiation fields. In this work, BANG-3 (MGS Research Inc.) gel vials from three production batches were irradiated with 6 MV photons of a Varian Clinac 2100 C linear accelerator and with the epithermal neutron beam of the Finnish boron neutron capture therapy (BNCT) facility at the FiR 1 nuclear reactor. The gel is tissue equivalent in main elemental composition and density and its T2 relaxation time is dependent on the absorbed dose. The T2 relaxation time map of the irradiated gel vials was measured with a 1.5 T magnetic resonance (MR) scanner using spin echo sequence. The absorbed doses of neutron irradiation were calculated using DORT computer code, and the accuracy of the calculational model was verified by measuring gamma ray dose rate with thermoluminescent dosimeters and 55Mn(n,γ) activation reaction rate with activation detectors. The response of the BANG-3 gel dosimeter for total absorbed dose in the neutron irradiation was linear, and the magnitude of the response relative to the response in the photon irradiation was observed to vary between different gel batches. The results support the potential of polymer gels in BNCT dosimetry, especially for the verification of two- or three-dimensional dose distributions.

Realistic neutron spectra for radiation protection and other applications at AERI, Budapest

The reconstruction of the Budapest Research Reactor (BRR) gave a good possibility to develop mixed neutron-gamma radiation fields for different applications like: simulation of operational spectra at power reactors, dosimeter development, neutron radiography, biological experiments. Recently, there are 3 horizontal channels available. In addition, isotopic neutron sources are in use in a separate laboratory. In a rotatable holder 4 different sources can be stored and automatically moved into irradiation position. There are changeable collimators and absorbers to modify the spectrum. In the large hall there are possibilities to study the room scatter, angular dependence of detectors, phantom albedo effect etc. Recently available sources are different Pu–Be (from 10 5–10 7 n/s yield), Ra–Be and Cf.

Radiation effect in silica optical fiber exposed to intense mixed neutron-gamma radiation field

We measured in situ the radiation-induced absorption of pure silica core fibers exposed to a fission nuclear reactor. We observed the growth of the 1.39-/spl mu/m. OH vibration band in polymer coated fiber. Three mechanisms are responsible for this effect: recoil protons, hydrogen diffusion, and a probable compaction effect. Based on this experiment, a fiber-optic neutron monitor prototype is proposed.

A new application of 3He and 10BF 3 proportional counters in a polythene moderator

A detailed examination of the pulse-height spectra from 3He and 10BF 3 proportional counters, placed inside a polythene moderator, and irradiated by neutrons, has led to the discovery of a “neutron event free-energy window”. It has been shown that this window can be used for the registration of gamma radiation without disturbing the registration of neutrons. A new method for measuring the ambient dose equivalent in mixed neutron-gamma radiation fields has been proposed.

Microdosimetric Characterisation of Radiation Fields with a High Pressure Ionisation Chamber: A Comparison with the Low Pressure Proportional Counter

Abstract The application is described of a tissue-equivalent (TE) high pressure ionisation chamber that permits evaluation of the mean quality factor in mixed neutron-gamma radiation fields and compares the data obtained with those acquired with a TE proportional counter (Rossi counter). Irradiations are produced by a research fission reactor. The radiation spectrum composition is modified by moderating the beam by various heavy water (D2O) or lead thicknesses and by inserting boral shutters in the neutron beam. Thus, radiation fields with most of the radiation dose originating from thermal neutrons or fast neutrons are obtained, with associated gamma ray absorbed doses ranging from 35% to 90%. The mean quality factors derived are in agreement (±10%) for situations where most of the radiation dose is due to fast neutrons. When most of the delivered dose is due to thermal or intermediate neutrons, the Rossi counter systematically predicts smaller (30 to 40%) quality factors than the high pressure TE chamber, this observation being confirmed on theoretical grounds, as most of the neutrons interacting with the Rossi counter do not cross the counter at these energies (i.e. they are 'stoppers'). The practical advantages and disadvantages of using either instrument in a particular situation are discussed.

The KFA Counter: A Dosimetry System for Use in Radiation Protection

Abstract The KFA counter, an area dose equivalent meter based on a low pressure tissue-equivalent proportional counter, was optimised with regard to its wall thickness to match H*(10) in the neutron energy range between thermal and 20 MeV. The measurements and calculations are summarised of the responses to monoenergetic neutrons and photons in terms of H*(10) and He(AP), the effective dose equivalent for anterior-posterior irradiation. Statistical precision as a function of neutron energy is also discussed. Measurements in unknown mixed neutron-gamma radiation fields near various work stations are presented.

The KFA Counter: A Dosimetry System for Use in Radiation Protection

The KFA counter, an area dose equivalent meter based on a low pressure tissue-equivalent proportional counter, was optimised with regard to its wall thickness to match H*(10) in the neutron energy range between thermal and 20 MeV. The measurements and calculations are summarised of the responses to monoenergetic neutrons and photons in terms of H*(10) and He(AP), the effective dose equivalent for anterior-posterior irradiation. Statistical precision as a function of neutron energy is also discussed. Measurements in unknown mixed neutron-gamma radiation fields near various work stations are presented.

Derivation of Radiation Quality Average Parameters in Neutron-γ Radiation Fields with the High-Pressure Ionization Chamber: Theory and Practice

The established radiation quality parameters in mixed neutron-gamma radiation fields may be measured by applying the initial (columnar) recombination of ions in tissue-equivalent (TE) high-pressure ionization chambers (recombination chambers). The mean quality factor can be determined to within 10-15% for mixed fields with neutrons ranging from thermal to 10 MeV, and the dose mean LET of the proton component can be determined to within 10-15% if the gamma-ray absorbed dose fraction is known. These average parameters are derived by measuring the ratio of the ionization currents collected at two high-field strengths and constant gas pressure applied to the ionization chamber. By utilizing approximate correlations between physical parameters in the neutron energy region from thermal to 10 MeV, the dose mean LET of the heavy ion component, the overall dose mean LET, and the microdosimetric parameter y0,D of the mixed field can also be derived. Experimental verification of the method is presented for various neutron-gamma radiation spectra in air and in water by comparison to theoretical calculations and results from low-pressure proportional counter measurements. Good agreement is shown. The TE high-pressure ionization chamber appears to have wide potential for use as a dose-equivalent meter in radiation protection or as a beam characterization device in radiobiology.

Measurements of Neutron Energy Using a Recoil-Proton Telescope and a High-Pressure Ionization Chamber

Two very different techniques for measuring the energy of neutrons in the energy range 0.1-10 MeV are presented and compared. A recoil-proton spectrometer is used to determine the energy spectra of neutrons produced by the d(4)-Be and p(4)-Be reactions down to the low-energy threshold of 0.7 MeV. The same radiation fields are also measured with a recently developed method using a high-pressure ionization chamber that can be used to determine the mean energy of the neutrons in a mixed neutron-gamma radiation field provided the gamma-ray absorbed dose fraction is determined independently. An intercomparison of the two methods shows that the high-pressure ionization chamber compares well and supplements the established recoil-proton spectrometer technique. The almost isotropic response of the chamber has enabled measurements to be made of the variation of mean neutron energy with depth in water for the two radiation fields.