Neutron Dose Equivalent Research Articles

Generation of neutron spectra and emission yields from proton-induced reactions on natC and natLi targets using CR-39 detectors

Fast neutron spectrometry and dosimetry at a specified location in a compact radiation environment is difficult due to large size of the conventional neutron spectrometers. In such situations, a small-size bare CR-39 detector can be used as a neutron detector to quantify the fast neutron fraction. In the present study, CR-39 detectors were placed very close to natural lithium and carbon targets during the proton irradiations to acquire the emission neutron spectra. The neutron spectra and the ambient dose equivalents in the forward and lateral directions with respect to the incident proton beam were estimated for both targets at incident proton energies between 8 and 20 MeV. The measured neutron spectra using CR-39 detectors could identify the quasi mono-energetic neutron features from the Li(p,n) and 13C(p,n) reactions effectively. An important observation of the present study is the identification of the fast neutron signature from the 13C(p,n) system from the discrete state de-excitations of excited 14N composite nuclei. The theoretical evaluation of the neutron spectral features and relative neutron energy distributions were performed using the FLUKA: FLUktuierende KAskade, a Monte Carlo simulation package, and the estimates agreed with the experimental results for both systems. The neutron ambient dose equivalents were also estimated from the measured spectra. These neutron fluence and dose estimates at close vicinity to the target can serve as an essential basis for shielding calculations and planning the pertinent radiation protection strategies.

Al–B4C-(Gd, Gd2O3) composite materials: Synthesis and characterization for neutron shielding applications

In this study, Al–20%B4C-x%Gd and Al–20%B4C-x%Gd2O3 (x = 1, 3, 5) composite powders were prepared using a high-energy planetary ball milling method to enhance the physical properties of Al–B4C neutron shielding composites. The prepared powders were subjected to uniaxial cold compaction at 500 MPa, resulting in cylindrical specimens. Subsequently, the specimens were sintered in a tube furnace at 600 °C for 1 h under an Ar atmosphere to prevent oxidation. The microstructure of the resulting composites was characterized using X-ray diffraction (XRD) and scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM/EDX). Archimedes density, hardness, and corrosion tests were performed on the compacted samples. Moreover, the composite's thermal and fast neutron absorption rates were calculated using the MCNP6.2 simulation code. The neutron equivalent dose rate was experimentally determined using the Am–Be neutron source. The simulation results demonstrated that the composite materials containing Gd exhibited the highest thermal neutron absorption rate, while those containing Gd2O3 demonstrated the highest fast neutron absorption rate. This research contributes valuable insights into the design and utilization of neutron-absorbing materials with suitable mechanical properties.

Photon and Neutron Dose Estimation Using Monte Carlo Simulation in TrueBeam’s Room

Purpose: The distribution of neutron ambient dose equivalent within the TrueBeam 10 MV photon chamber was investigated. Materials and Methods: The research used particle and heavy ion transport code system (PHITS) code and JENDL-5.0 to simulate the neutron ambient dose equivalent on and around TrueBeam’s head. The simulated results were compared with the measured results using CR-39 detectors when TrueBeam radiated 5000 monitor units of 10 MV photons with field sizes 20 cm × 20 cm and 0.5 cm × 0.5 cm. Results: Out of field size, the neutron ambient dose equivalents of the 0.5 cm × 0.5 cm field size are higher than those values of the 20 cm × 20 cm field size from 4% to 30%. The differences between the simulated value and the measured value of the neutron ambient dose equivalents at all points out of field size are smaller than 20%. Conclusion: The neutron ambient dose equivalents, simulated with PHITS and JENDL-5.0, are satisfied with the measured neutron ambient dose equivalent.

Dose equivalent consideration from neutron contamination in modified radiotherapy vault: a Monte Carlo study

Laminated barriers incorporating metal sheets provide effective protection for space-restricted radiotherapy centers. This study aimed to assess photoneutron contamination in smaller vaults protected by different compositions of multilayer barriers during simulated pelvic radiotherapy with 18 MV photon beams. Monte Carlo Simulations of 18 MV LINAC (Varian 2100 C/D) and Medical Internal Radiation Dose (MIRD) phantom were used to assess photoneutron contamination within reconstructed vaults incorporating different combinations of metal sheet and borated polyethylene (BPE) during pelvic radiotherapy. The findings highlight a 3.27 and 2.91 times increase in ambient neutron dose Hn* (10) along the maze of reconstructed vaults that use lead and steel sheets, respectively, compared to concrete. The Hn* (10) outside the treatment room increased after incorporating a metal sheet, but it remained within the permissible limit of 20 μSv/week for uncontrolled areas adjacent to the LINAC bunker, even with a workload of 1000Gy/week. Neutron equivalent doses in the patient’s organs ranged from 0.22 to 0.96 mSv Gy−1. There is no notable distinction in the organ’s neutron equivalent dose, fatal cancer risk, secondary radiation-induced cancer risk, and cancer mortality for various laminated barrier compositions. Furthermore, the use of metal sheets for vault wall reconstruction keeps the variation in cancer risk induced by photoneutrons below 6%, while risks of fatal cancer and cancer mortality vary less than 11%. While the metal portion of the laminated barrier raises the neutron dose, the addition of a BPE plate reduces concerns of increased effective dose and secondary malignancy risk.

Radiation shielding properties of composites of TiZrNbHfTa refractory high entropy alloy reinforced with TiZrNbHfTaOx high-entropy oxide

Radiation shielding materials play a critical role in advancing the next generation of nuclear reactors. Among these, high-entropy materials with refractory properties emerge as a new category for radiation shielding applications. This study introduces TiZrNbHfTa-TiZrNbHfTaOx composites tailored for such purposes. Initially, a TiZrNbHfTa refractory high-entropy alloy (RHEA) is synthesized via mechanical alloying, followed by the oxidation of the alloy to obtain its refractory high-entropy oxide (RHEO), TiZrNbHfTaOx. Composites are then prepared by blending 5%, 10%, and 20% weight fractions of RHEO with RHEA using mechanical alloying. Comparative analysis of shielding properties reveals a decrease in mass attenuation coefficient and effective atomic number with increasing RHEO content. While RHEA exhibits superior shielding performance against gamma photons and neutrons, the composite containing 20% RHEO emerges as the optimal shield against alpha and proton irradiations. The equivalent neutron dose rate and removal cross-section values are obtained 49.8% and 0.142 for the composite with 20% RHEO, respectively. These findings, supported by computational simulations in this study, contribute valuable insights for the advancement of novel shielding materials crucial for future nuclear reactor technologies.

Secondary neutron dosimetry for conformal FLASH proton therapy.

Cyclotron-based proton therapy systems utilize the highest proton energies to achieve an ultra-high dose rate (UHDR) for FLASH radiotherapy. The deep-penetrating range associated with this high energy can be modulated by inserting a uniform plate of proton-stopping material, known as a range shifter, in the beam path at the nozzle to bring the Bragg peak within the target while ensuring high proton transport efficiency for UHDR. Aluminum has been recently proposed as a range shifter material mainly due to its high compactness and its mechanical properties. A possible drawback lies in the fact that aluminum has a larger cross-sectionof producing secondary neutrons compared to conventional plastic range shifters. Accordingly, an increase in secondary neutron contamination was expected during the delivery of range-modulated FLASH proton therapy, potentially heightening neutron-induced carcinogenic risks to thepatient. We conducted neutron dosimetry using simulations and measurements to evaluate excess dose due to neutron exposure during UHDR proton irradiation with aluminum range shifters compared to plastic rangeshifters. Monte Carlo simulations in TOPAS were performed to investigate the secondary neutron production characteristics with aluminum range shifter during 225MeV single-spot proton irradiation. The computational results were validated against measurements with a pair of ionization chambers in an out-of-field region ( 30cm) and with a Proton Recoil Scintillator-Los Alamos rem meter in a far-out-of-field region (0.5-2.5m). The assessments were repeated with solid water slabs as a surrogate for the conventional range shifter material to evaluate the impact of aluminum on neutron yield. The results were compared with the International Electrotechnical Commission (IEC) standards to evaluate the clinical acceptance of the secondary neutronyield. For a range modulation up to 26cm in water, the maximum simulated and measured values of out-of-field secondary neutron dose equivalent per therapeutic dose with aluminum range shifter were found to be and , respectively, overall higher than the solid water cases (simulation: ; measurement: ). The maximum far out-of-field secondary neutron dose equivalent was found to be ( ) and ( ) for the simulations and rem meter measurements, respectively, also higher than the solid water counterparts (simulation: ( ) ; measurement: ( ) ). We conducted simulations and measurements of secondary neutron production under proton irradiation at FLASH energy with range shifters. We found that the secondary neutron yield increased when using aluminum range shifters compared to conventional materials while remaining well below the non-primary radiation limit constrained by the IECregulations.

Range shifter contribution to neutron exposure of patients undergoing proton pencil beam scanning.

Superficial targets require the use of the lowest energies within the available energy range in proton pencil-beam scanning (PBS) technique. However, the lower efficiency of the energy selection system at these energies and the requirement of a greater number of layers may represent disadvantages for this approach. The alternative is to use a range shifter (RS) at nozzle exit. However, one of the concerns of using this beamline element is that it becomes an additional source of neutrons that could irradiate organs situated far from the target. The purpose of this study is to assess the increase in neutron dose due to the RS in proton PBS technique. Additionally, an analytical model for the neutron production is tested. Two clinical plans, designed to achieve identical target coverage, were created for an anthropomorphic phantom. These plans consisted of a lateral field delivering an absorbed dose of 60Gy (RBE) to the target. One of the plans employed the RS. The MCNP code was used to simulate the plans, evaluating the distribution of neutron dose equivalent (Hn) and the equivalent dose in organ. In the plan with the RS plan, neutron production from both the patient and the RS were assessed separately. Hn values were also fitted versus the distance to field edge using a Gaussian function. Hn per prescription dose, in the plan using the RS, ranged between 1.4 and 3.7 mSv/Gy at the field edge, whereas doses at 40cm from the edge ranged from 9.9 to 32 μSv/Gy. These values are 1.2 to 10 times higher compared to those obtained without the RS. Both this factor and the contribution of neutrons originating from the RS increases with the distance from field edge. A triple-Gaussian function was able to reproduce the equivalent dose in organs within a factor of 2, although underestimating the values. The dose deposited in the patient by the neutrons originating from the RS predominantly affects areas away from the target (beyond approximately 25cm from field edge), resulting in a neutron dose equivalent of the order of mSv. This indicates an overall low neutron contribution from the use of RS in PBS.

The influence of Monte Carlo radiation transport codes and hadron physics models on stray neutron dose estimations in proton therapy: An extended intercomparison

PurposeSelecting a Monte Carlo (MC) radiation transport code and nuclear physics model affects the simulation of secondary neutron fluence energy spectra and the estimation of the neutron dose equivalent in proton therapy. This study compares the MC codes FLUKA, MCNPX and GEANT4 from previous work with two additional MC codes, TOPAS (with either the Binary Intra-nuclear Cascade (BIC) or the Bertini intra-nuclear cascade model (Bert)) and PHITS (with either the Intra-Nuclear Cascade of Liege (INCL) or the Bert model), to facilitate a more comprehensive comparison. Material and methodsSecondary neutron fluence energy spectra were scored at different locations inside a water phantom, exposed to a 10 × 10 cm2 parallel mono-energetic proton beam with energies of 110 MeV, 150 MeV, 180 MeV and 210 MeV. The neutron dose equivalent was estimated by applying fluence-to-dose equivalent conversion coefficients. ResultsDiscrepancies between MC codes and physics models reach 90 % for the total neutron fluence. Differences are most pronounced in the forward direction of the beam due to a notable presence of high-energy neutrons and the higher uncertainty on the neutron cross-section data for these neutrons. Nuclear model selection within PHITS (INCL or Bert) and TOPAS (BIC or Bert) shows a worse agreement for the Bert model than for the other MC codes and nuclear models. The discrepancies in the neutron spectrum simulations propagate into differences in the estimated neutron dose equivalents. They are more prominent for distal positions (up to a factor of 3) than for lateral positions (up to a factor of 2) and decrease at higher proton energies. ConclusionsThis study shows considerable differences in the out-of-field neutron fluence energy spectra calculated by different MC codes and nuclear physics models. Differences in the neutron dose equivalent reach a factor of 3. The discrepancies are most apparent in distal positions and at lower proton energies.

Measurement of energy and directional distribution of neutron ambient dose equivalent for the [formula omitted]Li(p,n)[formula omitted]Be reaction

Double differential neutron fluence distributions were measured in the 7Li(p,n)7Be reaction for proton beam energies 7, 9 and 12 MeV. Seven liquid scintillator based detectors were employed to measure neutron fluence distributions using the Time of Flight technique. Neutron ambient dose equivalents were determined from the measured fluence distribution using ICRP (International Commission on Radiological Protection) recommended fluence to dose equivalent conversion coefficients. Neutron dose equivalents were also measured using a conventional BF3 detector based REM counter. Ambient dose equivalent measured by the REM counter is found to be in agreement with that determined from the neutron fluence spectra within their uncertainties. Angular distributions of the ambient dose equivalents were also determined from the measured fluence distributions at different angles.

Thermal neutron detection and track recognition method in reference and out-of-field radiotherapy FLASH electron fields using Timepix3 detectors.

Objective.This work presents a method for enhanced detection, imaging, and measurement of the thermal neutron flux.Approach. Measurements were performed in a water tank, while the detector is positioned out-of-field of a 20 MeV ultra-high pulse dose rate electron beam. A semiconductor pixel detector Timepix3 with a silicon sensor partially covered by a6LiF neutron converter was used to measure the flux, spatial, and time characteristics of the neutron field. To provide absolute measurements of thermal neutron flux, the detection efficiency calibration of the detectors was performed in a reference thermal neutron field. Neutron signals are recognized and discriminated against other particles such as gamma rays and x-rays. This is achieved by the resolving power of the pixel detector using machine learning algorithms and high-resolution pattern recognition analysis of the high-energy tracks created by thermal neutron interactions in the converter.Main results. The resulting thermal neutrons equivalent dose was obtained using conversion factor (2.13(10) pSv·cm2) from thermal neutron fluence to thermal neutron equivalent dose obtained by Monte Carlo simulations. The calibrated detectors were used to characterize scattered radiation created by electron beams. The results at 12.0 cm depth in the beam axis inside of the water for a delivered dose per pulse of 1.85 Gy (pulse length of 2.4μs) at the reference depth, showed a contribution of flux of 4.07(8) × 103particles·cm-2·s-1and equivalent dose of 1.73(3) nSv per pulse, which is lower by ∼9 orders of magnitude than the delivered dose.Significance. The presented methodology for in-water measurements and identification of characteristic thermal neutrons tracks serves for the selective quantification of equivalent dose made by thermal neutrons in out-of-field particle therapy.

Measurement of undesirable neutron spectrum in a 120 MeV linac

Photoneutron background spectroscopy observations at linac are essential for directing accelerator shielding and subtracting background signals. Therefore, we constructed a Bonner Sphere Spectrometer (BSS) system based on an array of BF3 gas proportional counter tubes. Initially, the response of the BSS system was simulated using the MCNP5 code. Next, the response of the system was calibrated by using neutrons with energies of 2.86 MeV and 14.84 MeV. Then, the system was employed to measure the spectrum of the 241Am–Be neutron source, and the results were unfolded by using the Gravel and EM algorithms. Using the validated system, the undesirable neutron spectrum of the 120 MeV electron linac was finally measured and acquired. In addition, it is demonstrated that the equivalent undesirable neutron dose at a distance of 3.2 m from the linac is 19.7 μSv/h. The results measured by the above methods could provide guidance for linac-related research.

Results from the Radiation Assessment Detector on the International Space Station: Part 3, combined results from the CPD and FND

The energetic particle radiation environment on the International Space Station (ISS) includes both charged and neutral particles. Here, we make use of the unique capabilities of the Radiation Assessment Detector (ISS-RAD) to measure both of these components simultaneously. The Charged Particle Detector (CPD) is, despite its name, capable of measuring neutrons in the energy range from about 4 MeV to a few hundred MeV. Combined with data from the Fast Neutron Detector (FND) in the 0.2 to 8 MeV range, we present the first broad-spectrum measurements of the neutron environments in various locations within the ISS since an early Bonner-Ball experiment that was conducted before the Station was fully constructed. The data presented here span the time period from February 2016 to February 2022. In addition to presenting broad-spectrum neutron fluence measurements, we show correlations of the measured neutron dose equivalent with charged-particle dose rates. The ratio of charged-particle dose to neutron dose equivalent is found to be relatively stable within the ISS.

Impact of supporting substrate on the measured neutron yields from natLi (p, n) system: A case study with natC and natTa stopping targets

Abstract The high energy neutron generation facilities worldwide uses the Li(p,n) system for production of the quasi mono-energetic neutrons from epithermal to few hundreds of MeV. Among the facilities either self-standing thick Li target or Li foil supported on a target assembly is used. In case of a supported Li target, either low or high mass supporting substrates like Al, C, Au, Ta etc. can be used. However, the neutron emission from the supporting substrate can interfere with the Li(p,n) reaction neutrons. In the present study neutron yields from a thin natural Li target supported on the thick natural Ta (high Z) and C (low Z) targets were measured at incident proton energies between 8-20 MeV. The proton irradiation of targets were performed at the Pelletron accelerator Facility, Mumbai and the emitted neutrons were measures at both forward (0°) and lateral (90°) directions with respect to the incident beam. The neutron yield was estimated using the CR-39 detectors and dual scintillator based active neutron spectrometer. A neutron dose equivalent (NDE) meter was also used to measure the neutron ambient dose equivalent at the measurement point. Experimentally measured neutron yields and ambient neutron dose equivalent estimates indicated higher Li(p,n) neutron yield with carbon as the supporting substrate. This is due to fewer self-emissions from natC compared to natTa target. Carbon supporting substrate also ensures improved spectral features of the Li(p,n) emission neutrons and lesser neutron fluence around the source location. This ensures efficient implementation of radiation protection practices with carbon target due to lower neutron fluence around the source compared to tantalum target. Carbon is a favorable supporting substrate compared to high Z- metal targets like Ta for emission neutron studies on Li target at proton energies between 8-20 MeV.

Dosimetric effects of different hip prosthesis materials during pelvic radiotherapy using high energy photons

This study investigates the effects of different hip prosthesis materials on the secondary neutron and electron contamination produced by high energy, i.e. 18 MV, photon beams during pelvic irradiation. Three different materials, including Titanium-alloy (Ti-alloy), Stainless Steel (SST), and Cobalt–Chromium–Molybdenum-alloy (Co–Cr–Mo), were evaluated in a Monte Carlo (MC) study. The Varian 2100 C/D linac head and an ICRP reference adult male voxel phantom with a realistic hip prosthesis were simulated using the MCNPX (ver. 2.6) MC code. As expected, results showed that secondary electron and neutron generation mainly depends on the shape and material type of the prosthesis. The mean dose increase factor (DIF) for electron contamination is 1.72, 1.53, and 1.35 for Co–Cr–Mo, SST, and Ti-alloy, respectively, at the tissue-prosthesis interface. Electron contamination also depends on the prosthesis thickness and it is increased by increasing the thickness. The maximum DIF for neutrons occurs at the center of the prosthesis stem and its mean DIF is 2.91, 2.49, and 2.34 underneath Co–Cr–Mo, SST, and Ti-alloy prostheses, respectively. The amount of neutron DIF is slightly reduced above prostheses. Furthermore, the results indicate that the most significant change in the electron dose distribution is observed at the tissue-prosthesis interface. The neutron dose equivalent is increased from 76% to 200% in the presence of the different materials of the prosthesis. Hip prostheses made of Ti-alloy can significantly decrease the dose due to neutron and electron contamination, compared to the other materials considered in the simulations.

Dosimetric evaluation in brachytherapy and teletherapy for prostate cancer using Monte Carlo simulation and anatomically realistic virtual anthropomorphic phantom

The benefits that radiotherapy provides to the patient are unquestionable. However, there is growing concern about the increased risk of radiation-induced secondary cancer and the late damage to organs and tissues associated with the treatment. In this sense, the main objective of this study was to determine the radiation doses in the organs and tissues of the prostate cancer patient for two radiotherapy options, using the Monte Carlo code MCNP×2.7.0. The patient was represented by the ICRP 110 reference virtual anthropomorphic phantom and underwent brachytherapy with a192Ir source, Nucletron Microselectron V2 model, and alternatively, conventional 3D-CRT radiotherapy with an 18 MV X-ray beam from a Varian 2100c medical linear accelerator with 4-field planning (anteroposterior, 0°; posteroanterior, 180°; left lateral, 90°; and right lateral, 270°). A set of conversion factors (CF) for equivalent and effective dose were determined. The obtained results showed that in the more distant organs related to the prostate, such as the lung, brain, and salivary glands, the contribution of scattered radiation decreases, especially when brachytherapy is used. Comparing the two treatment modalities, the CF for effective dose per therapeutic dose was 42% lower when 192Ir was chosen as the main form of treatment. The CF values for photon equivalent doses calculated in 3D-CRT 4-fields are among one and two orders of magnitude higher than 192Ir therapy. These differences are even higher when the neutron equivalent dose in 3D-CRT is considered. In this way, the tools and methodology presented significantly improve the previously used to accurately estimate the equivalent doses in organs and tissues, which is the best quantity to assess the risk of inducing secondary cancer of healthy organs and tissues of patients undergoing treatment with radiotherapy.

FOIL ACTIVATION TECHNIQUE-A TOOL FOR THE EVALUATION OF PHOTO-NEUTRON DOSE IN RADIOTHERAPY.

Treatment of cancer is carried out using photon beams from high-energy medical linear accelerators. Photo-neutrons are also produced as an unwanted by product in the process of dose delivery to the cancer patients during their radiation treatments. In the present study, photo-neutron dose equivalents (both thermal and fast components) per unit delivered gamma-photon dose were measured at different depths, as function of distances from iso-centre in patient plane, field sizes, wedge angles and at LINAC head for a 15-MV medical linear accelerator model Elekta Precise using multi-foil activation technique. The neutron dose equivalents determined for the above-mentioned parameters were found to be lower (<0.05%) in comparison with the therapeutic photon dose delivered and within the prescribed limits recommended by the national regulatory authority.

Comparison of neutron contamination in small photon fields of secondary collimator jaws and circular cones

Abstract Introduction: Advanced treatment modalities involve applying small fields which might be shaped by collimators or circular cones. In these techniques, high-energy photons produce unwanted neutrons. Therefore, it is necessary to know neutron parameters in these techniques. Materials and methods: Different parts of Varian linac were simulated by MCNPX, and different neutron parameters were calculated. The results were then compared to photoneutron production in the same nominal fields created by circular cones. Results: Maximum neutron fluence for 1 × 1, 2 × 2, 3 × 3 cm2 field sizes was 165, 40.4, 19.78 (cm–2.Gy-1 × 106), respectively. The maximum values of neutron equivalent doses were 17.1, 4.65, 2.44 (mSv/Gy of photon dose) for 1 × 1, 2 × 2, 3 × 3 cm2 field size, respectively, and maximum neutron absorbed doses reached 903, 253, 131 (µGy/Gy photon dose) for 1 × 1, 2 × 2, 3 × 3 cm2 field sizes, respectively. Conclusion: Comparing the results with those in the presence of circular cones showed that circular cones reduce photoneutron production for the same nominal field sizes.

Effect of the B4C content on microstructure, microhardness, corrosion, and neutron shielding properties of Al–B4C composites

Aluminum–Boron Carbide (Al–B4C) metal matrix composites (MMCs) are frequently studied because of their high thermal conductivity and ability to absorb neutrons, as an alternative material for the manufacture of baskets in spent fuel casks in the nuclear industry. Keeping in mind, this study aimed to produce Al–B4C MMCs and examine their properties to shield against neutron radiation. The Al-xB4C (x = 5, 10, 15, 20, 25, 30, 40, 45, and 50 wt%) composite powders were prepared by a high-energy planetary ball milling. The prepared powders were compacted with uniaxial cold compaction at 500 MPa and sintered in a tube furnace at 600 °C for 1 h under an Argon atmosphere. The microstructure of the obtained MMCs was characterized by using X-ray diffraction (XRD) and scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM/EDX). The XRD result indicates that the synthesized Al–B4C MMCs have a predominant phase of Al and B4C. The SEM images show the distribution of B4C reinforcement inside the Al matrix. The relative density, measured by the Archimedes principle decreased from 99.95 to 91.55% as the B4C ratio increased. The microhardness value was increased in the beginning from 44 to 75.5 HV and then decreased to 36.5 HV as the B4C ratio increased. The polarization curves show that the corrosion current density and the corrosion rate were significantly increased from 52.25 to 375.36 μA cm−2 and 0.573–4.121 mm/year as the B4C ratio increased. The simulation results, computed by the MCNP6.2 program show that the thermal neutron cross-section and the fast neutron cross-section was increased from 4.5 to 40.1 cm−1 and 0.13 to 0.15 cm−1 as the B4C ratio increased. Further, the neutron equivalent dose rate was calculated experimentally with the 241Am–Be neutron source and the data was completely agreed with the computational results.

Neutron area monitor for ambient dose equivalent (H*(10)) and effective dose (EAP) using prompt gamma rays induced in high density polyethylene with metallic external layer

Estimations of neutron ambient dose equivalent, H∗(10) and effective dose (anterior-posterior), EAP using measured prompt gamma intensities are theoretically investigated for several prompt gamma generating detection systems, each consisting of a cylindrical high density polyethylene (HDPE) covered with a layer of different materials. The neutron induced prompt gamma intensities are measured using a NaI(Tl) detector with lead as shielding. Materials like cadmium, iron, nickel, zinc, copper and lead are used as an outer layer covering the HDPE cylinder. A linear combination of the distributions of the prompt gamma intensities emitted from these materials at different incident neutron energies are subjected to multiple linear regression analyses to fit the energy dependent neutron fluence to dose conversion coefficients (DCC) as provided by the International Commission on Radiological Protection (ICRP). The aim of the present work is to explore the suitability of other elements to replace boron in the earlier used borated HDPE in the system to overcome certain practical difficulties. Different neutron energy distributions in the energy range from about 0.01 eV to 20 MeV are selected to validate the dose estimation. Among the materials used, iron, nickel, zinc and copper show promising results for the estimation H∗(10) as well as EAP as a replacement of boron for neutron energy distributions with predominant high energy components.

Experimental and Monte Carlo based study of different microdosimetric quantities at mixed radiation environments of nuclear reactor, reprocessing facility and D-D accelerator

Personnel working in the environment of nuclear reactors, fuel reprocessing facilities, high energy accelerators get exposed to a mixed radiation field of electrons, photons, and neutrons with a wide range of energy and fluence. Performance evaluations of the passive personnel dosimeters in these mixed radiation environments require knowledge of the external dose contributed by different types of radiations having variation in their energy, etc. Tissue Equivalent Proportional Counters (TEPC), sensitive to a wide range of particles and energy (Linear Energy Transfer (LET) ∼ 0–1500 keV μm−1), is one of the standard reference detectors which can be used for the characterization of mixed radiation fields. In the present work, HAWK TEPC was used to evaluate different microdosimetric parameters such as dose mean lineal energy (ӯD), average quality factor (Ǭ), the contribution of gamma and neutron dose equivalents for the characterizations of the radiation fields inside in a nuclear research reactor hall, fuel reprocessing facility, D-D neutron beam facility, radionuclide source based neutron calibration facilities, etc. Measured results were compared with Monte Carlo simulated values using radiation transport code PHITS. TEPC measurement shows that the radiation work field around nuclear reactor facility is dominated by gamma rays with approximate ӯD and Ǭ of ∼42 keV μm−1 and ∼2.3, respectively. Whereas ӯD and Ǭ values for neutron dominated locations in fuel reprocessing facility are 82 keV μm−1 and 8.2, respectively. Ǭ values near thermal neutron beam line in research reactor are close to 1 due to (n, γ) reaction. For 2.54 MeV mono-energetic D-D neutron beam, measured and calculated dose mean lineal energies (ӯD) are 52.5 keV μm−1 and 59.6 keV μm−1, respectively, whereas corresponding Ǭ values are 11.23 and 13.01. Obtained values can be utilized to correlate the neutron and gamma dose equivalents measured by passive detectors used for personnel monitoring program to enhance confidence in the dosimetry.