Monte Carlo simulation of scattered and thermal photoneutron fluences inside a radiotherapy room

In this work, the behaviour of scattered and thermal photoneutron fluences in a radiotherapy treatment room was investigated by means of Monte Carlo simulation. The MCNP code was used to study the dependence of these neutron fluences on room design, room area and the effect of neutron moderator materials placed on the room walls. The results of the investigation showed a poor agreement between the simulations and empirical approximations, suggesting that the formulae found in the literature can underestimate the neutron flux inside a radiotherapy room.     

Detection unit optimization of a neutron searching detector using Monte Carlo Simulations

The NSD is a portable Neutron Searching Detector developed at Rotem Industries Ltd. with a high efficiency for counting fast and thermal neutrons employing improved gamma rejection. The NSD detection-unit consists of two ³He detectors installed within a polypropylene moderator. The latest international standards for detection of illicit trafficking of radioactive materials require high sensitivity, relatively small dimensions, and light mass. In order for it to meet these standards, the NSD detection-unit was optimized using Monte Carlo N-Particle transport code (MCNP). The moderator mass and dimensions were reduced without deterioration, even improving the instrument's sensitivity. The purposed moderator improvements covered in this paper work well for traditional hand-held neutron search detectors based on ³He tubes as well as for new neutron detection technologies due to the severe worldwide shortage of ³He.Three geometrical moderator configurations were examined using the MCNP code—a rectangular box, a circular cylinder, and an elliptical base cylinder. The optimization results showed that both the rectangular box moderator and the elliptical base cylinder moderators achieve the appropriate sensitivity required by the standards with about 30% reduced mass. A prototype was fabricated with the rectangular box moderator configuration, and its response was successfully validated by comparing empirical measurements against the results of the MCNP code.Performance examination of the optimal detection unit prototype was made regarding the latest international standards. The results showed a 17% improvement in detection limit for radioactive materials along with a 14% to 17% increased neutron detection response, while keeping the false alarm rate below the required threshold, and maintaining a 26% mass reduction.     

Monte Carlo calculation of the response function for a He neutron spectrometer

The response of a gridded ³He ionization chamber to monoenergetic neutrons has been calculated using a Monte Carlo approach. The effects of neutron scattering on detector materials, wall effects, recoil continua and background neutrons are included. The calculated results are smoothed according to a peak shape taken from experiment. Response functions in the energy range E_n < 1217 keV are compared to experimental data obtained with the ⁷Li(p, n)⁷Be reaction and to the detector efficiency derived from them.     

Thermoluminescence dosimetry of a thermal neutron field and comparison with Monte Carlo calculations

The characteristics of thermoluminescence dosemeters (TLDs) regarding the determination of photon and neutron absorbed doses were investigated in a thermal neutron beam. Harshaw TLD-100 (LiF:Mg,Ti) and TLD-700 (7LiF:Mg,Ti) were compared with similar materials from Solid Dosimetric Detector and Method Laboratory (People's Republic of China). Harshaw TLD-700H (7LiF:Mg,Cu,P) and aluminium oxide (Al2O3:Mg,Y) from Hungary were also considered for photon dose measurement. The neutron sensitivity of the investigated materials was measured and found to be consistent with values reported by other authors. A comparison was made between the TL dose measurements and results obtained via conventional methods. An agreement within 20% was obtained, which demonstrates the ability of TLD for measuring neutron and photon doses in a mixed field, using careful calibration procedures and determining the neutron sensitivity for the usage conditions.     

Monte Carlo simulation of grating-based neutron phase contrast imaging at CPHS

Since the launching of the Compact Pulsed Hadron Source (CPHS) project of Tsinghua University in 2009, works have begun on the design and engineering of an imaging/radiography instrument for the neutron source provided by CPHS. The instrument will perform basic tasks such as transmission imaging and computerized tomography. Additionally, we include in the design the utilization of coded-aperture and grating-based phase contrast methodology, as well as the options of prompt gamma-ray analysis and neutron-energy selective imaging. Previously, we had implemented the hardware and data-analysis software for grating-based X-ray phase contrast imaging. Here, we investigate Geant4-based Monte Carlo simulations of neutron refraction phenomena and then model the grating-based neutron phase contrast imaging system according to the classic-optics-based method. The simulated experimental results of the retrieving phase shift gradient information by five-step phase-stepping approach indicate the feasibility of grating-based neutron phase contrast imaging as an option for the cold neutron imaging instrument at the CPHS.     

Implementation of variance-reduction techniques for Monte Carlo nuclear logging calculations with neutron sources

Monte Carlo simulations for nuclear logging applications are considered to be highly demanding transport problems. In this paper, the implementation of weight-window variance reduction schemes in a 'manual' fashion to improve the efficiency of calculations for a neutron logging tool is presented. The simulations were conducted using MCNP to assess the sensitivity of neutron and gamma flux to the saturation of soil with fresh and salty water and oil. The weight-windows were optimised to enhance the precision in characteristic gamma-ray windows for carbon-oxygen logging and sigma logging. An analysis is under way to incorporate quartz sediments and other oil substances. Future tasks include the assessment of advanced Monte Carlo estimators and optimisation of capture gamma-ray sampling techniques to enhance the variance reduction in nuclear logging calculations.     

Three-dimensional shielding calculations for the IFMIF neutron source using a coupled Monte Carlo/Deterministic computational scheme

Shielding calculations for the International Fusion Materials Irradiation Facility (IFMIF) are complicated due to the geometrical complexity of the target system and the large-scale bulk shields around the source target. Three-dimensional shielding calculations were performed by using a newly developed Monte Carlo/Deterministic computational scheme. The neutron-photon fluxes and dose rate distributions in the back wall of the Test Cell and the access/maintenance room are presented and compared with previous shielding calculations. The results demonstrate that this coupled scheme is an useful computational tool for three-dimensional shielding analyses of complex and large nuclear facilities.     

Field parameters and dosimetric characteristics of a fast neutron calibration facility: experimental and Monte Carlo evaluations

At the ENEA Institute for Radiation Protection (IRP) the fast neutron calibration facility consists of a remote control device which allows locating different sources (Am–Be, Pu–Li, bare and D₂O moderated ²⁵²Cf) at the reference position, at the desired height from the floor, inside a 10×10×3 m³ irradiation room. Either the ISO reference sources or the Pu–Li source have been characterised in terms of uncollided H^*(10) and neutron fluence according to the ISO calibration procedures. A spectral fluence mapping, carried out with the Monte Carlo Code MCNP^TM, allowed characterising the calibration point, in scattered field conditions, according to the most recent international recommendations. Moreover, the irradiation of personal dosemeters on the ISO water filled slab phantom was simulated to determine the field homogeneity of the calibration area and the variability of the neutron field (including the backscattered component) along the phantom surface. At the ENEA Institute for Radiation Protection the calibration of neutron area monitors as well as personal dosemeters can now be performed according to the international standards, at the same time guaranteeing suitable conditions for research and qualification purposes in the field of neutron dosimetry.     

Assessment of sensitivities and uncertainties in Monte Carlo particle transport calculations for neutron spectrometry

Monte Carlo particle transport (MCPT) calculations are increasingly used in neutron metrology and especially in neutron spectrometry for the numerical simulation of measurements based on computational models for the underlying experiment. Sensitivities express the dependence of results on input data and uncertainties are a measure of the dispersion of calculated or measured values. The knowledge of sensitivities is instrumental for physical understanding and a possible optimisation of computational models and/or the experiments. The assessment of uncertainties is necessary to state the degree of belief in results and to objectively validate calculations by measurements. Methods used in MCPT calculations which assess sensitivities of the computational model and propagate uncertainties associated with input data are discussed. The technique most often used is demonstrated by a simple example.     

Standardisation of water-moderated 241Am-Be neutron source using De Pangher neutron long counter: experimental and Monte Carlo modelling

A convenient neutron source is made for calibration of neutron survey instruments and personal dosimeters that are used in various nuclear installations such as fuel reprocessing, waste management, fuel fabrication and oil and well logging facilities, etc. This source consists of a bare (241)Am-Be neutron source placed at the centre of a 15-cm radius stainless steel spherical shell filled with distilled water. This paper describes the standardisation of the source at Bhabha Atomic Research Centre, using De Pangher neutron long counter both experimentally and using the Monte Carlo simulation. The ratio of neutron yield of water moderated to the bare (241)Am-Be neutron source was found to be 0.573. From the simulation, the neutron-fluence-weighted average energy of water-moderated (241)Am-Be source (fluence-weighted average energy of 2.25 MeV, dose-weighted average energy of 3.55 MeV) was found to be nearly the same as that of a (252)Cf source (fluence-weighted average energy of 2.1 MeV, dose-weighted average energy of 2.3 MeV). This source can be used for calibration in addition to (252)Cf, to study the variation in response of neutron monitoring instruments.     

Calculated organ equivalent doses for individuals in a sitting posture above a contaminated ground and a PET imaging room

In this paper, phantoms representing sitting postures were developed and implemented in the MCNPX code to perform dose calculations. For the ground contamination case, isotropic planar source of (137)Cs was used. The male sitting phantom received an effective dose rate of 37.73 nSv h(-1) for a contamination of 30 kBq m(-2). The gonadal equivalent dose of the male sitting phantom was 40 % larger than that from the standing phantom. For the positron emission tomography clinic, a point photon source with the energy of 511 keV was used. The gonadal equivalent dose of the male sitting phantom was 117 % larger than that for the standing phantom. For an 8-h day, the effective dose of the sitting phantom was 2.54 μSv for 550 MBq F-18. This study concludes that phantoms with realistic postures could and should be considered in organ equivalent dose calculations in certain environmental and medical dosimetry studies where accurate data are desired.     

Adjoint acceleration of Monte Carlo simulations using TORT/MCNP coupling approach: a case study on the shielding improvement for the cyclotron room of the Buddhist Tzu Chi General Hospital

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the original inadequate maze design. Variance reduction techniques are indispensable in this study to facilitate the simulations for testing a variety of configurations of shielding modification. The TORT/MCNP manual coupling approach based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology has been used throughout this study. The CADIS utilises the source and transport biasing in a consistent manner. With this method, the computational efficiency was increased significantly by more than two orders of magnitude and the statistical convergence was also improved compared to the unbiased Monte Carlo run. This paper describes the shielding problem encountered, the procedure for coupling the TORT and MCNP codes to accelerate the calculations and the calculation results for the original and improved shielding designs. In order to verify the calculation results and seek additional accelerations, sensitivity studies on the space-dependent and energy-dependent parameters were also conducted.     

Evaluation of N50 neutron slab monitor detection efficiency with TRIPOLI-4.3 Monte Carlo code

Monte Carlo code TRIPOLI-4.3 has been used to evaluate the N50 neutron slab monitor detection efficiency benchmark organised by the non-destructive assay (NDA) working group of European Safeguards Research and Development Association (ESARDA). The lattice geometry of TRIPOLI-4.3 was applied to simplify the modelling of the N50 monitor, which consists of four 3He tubes embedded in a high-density polyethylene (HDPE) moderator. Using additional HDPE slabs and/or Cd sheets, 10 geometry configurations with different states of moderation were considered in this study. Two cross section libraries from ENDF/B-VI.4 and JEF2.2 and three representations of the 252Cf fission spectrum were investigated and the calculation experiment (C/E) ratios were discussed. The perturbation option was used in order to study the sensitivity to the HDPE density. The S(alpha,beta) thermal scattering in HDPE was also tested to quantify its impact on the N50 detection efficiency calculations.     

Neutron spectrometry in a PET cyclotron with a Bonner sphere system

Positron emission tomography (PET) is a non-invasive medical imaging technique normally used for diagnostic purposes to determine the location and concentration of physiologically active compounds in a human body. An unshielded cyclotron is used for PET at the Clinica Universitaria de Navarra to produce short-lived positron emitting radionuclides ((15)O, (13)N, (11)C and (18)F) by bombarding appropriate target material with proton or deuteron beams with energies up to 18 and 9 MeV, respectively. Subsequent nuclear reactions may generate undesirable neutrons that should be evaluated and controlled. In this study, the neutron measurements performed with an active and a passive Bonner sphere systems at different locations outside and inside the cyclotron vault during operation have been presented. The neutron spectrum at each location was determined with an unfolding code developed by the authors.     

Monte Carlo calculation of ultra cold neutron flow through long tubes with a realistic angular distribution of reflected neutrons

Monte Carlo simulations of the flow of particles through tubes are used to study the effects of a theoretically well founded non-specular reflection probability on the transmission of the tubes and the angular distribution of the emerging particles. The theoretical reflection probability used is appropriate for ultra cold neutrons scattering from almost smooth surfaces. The results obtained are compared to those described in earlier work in which approximations to the theoretical reflection law were used.     

Neutron measurements in the vicinity of a self-shielded PET cyclotron

The radionuclides used in positron emission tomography (PET) are short-lived and generally must be produced on site using a cyclotron. A common end product of the nuclear reactions used to produce the PET radionuclides is neutron radiation. These neutrons could potentially contribute to the annual effective dose received by hospital personnel. A Bonner sphere spectrometer was used to measure neutron energy spectra at three locations near a self-shielded PET cyclotron. This cyclotron accelerates protons to 11 MeV. The neutron measurements reported were made during the production of 18F via the 18O(p,n)18F reaction (Q = -2.4 MeV). Neutron spectra were obtained with the BUMS unfolding code and converted to dose equivalent rates.     

Monte Carlo modelling of a simple accident dosemeter

A simple dosemeter made of a sulphur tablet, bare and cadmium-covered indium foils and a cadmium-covered copper foil has been modelled using MCNP5. Studies of the model without phantom or other confounding factors have shown that the cross sections and fluence-to-dose factors generated by the Monte Carlo method agree with those generated by analytic expressions for the high energy component. In this study, a comparison of the effect of location on phantoms and an extension to low and intermediate energies is done. The activities expected from exposure to four critical assemblies on phantom is calculated and compared with observations.     

Magnetism in Nanoislands: a Monte Carlo Study

The properties of nanoislands consisting of mixed spins-2 and -5/2 have been investigated using Monte Carlo simulations. Some interesting phenomena have been found in the thermal variations of the nanoisland system. The shapes of total magnetization and total magnetic susceptibility are greatly influenced by the exchange couplings. The reduced transition temperature of the nanoisland is deduced. The magnetic hysteresis cycles are obtained for different temperatures, crystal fields, and exchange interactions. Some results obtained in the nanoisland may have potential applications in different research fields, such as electronics, optics, mechanics, and even biomedicine and molecular devices.     

Monte Carlo simulation of a transition radiation detector

This paper employs Monte Carlo simulations of the performance of a transition radiation detector (TRD). The program has been written for the TRD in the ZEUS spectrometer, which separates electrons from hadrons in the momentum range between 1 and 30 GeV/c. Both, total charge method and cluster counting method were simulated taking into account various experimental parameters. In particular, it was found that the cluster counting method relies on a quantitative understanding of the background originating from the production of δ-electrons by charged particles. The results of the Monte Carlo calculations are in agreement with experimental data obtained with prototypes within a systematic uncertainty of 20%. We applied our Monte Carlo program to studies in order to find an optimum layout for the TRD within available space in the ZEUS spectrometer. In this context, the performance of TRD layouts with different geometries and materials has been evaluated comprehensively. The geometry found by optimization promises an improvement on hadron suppression by a factor of about two for both methods compared with present results from test measurements. Applying algorithms for a detailed analysis of the energy and space distributions of the clusters in the TRD, hadrons in the momentum range from 1 to 30 GeV/c can be suppressed to a level of less than 2%. This method of cluster analysing improves the suppression of hadrons by a factor of about two compared to the total charge method.     

Monte Carlo simulation of a mammographic test phantom

A test phantom, including a wide range of mammographic tissue equivalent materials and test details, was imaged on a digital mammographic system. In order to quantify the effect of scatter on the contrast obtained for the test details, calculations of the scatter-to-primary ratio (S/P) have been made using a Monte Carlo simulation of the digital mammographic imaging chain, grid and test phantom. The results show that the S/P values corresponding to the imaging conditions used were in the range 0.084-0.126. Calculated and measured pixel values in different regions of the image were compared as a validation of the model and showed excellent agreement. The results indicate the potential of Monte Carlo methods in the image quality-patient dose process optimisation, especially in the assessment of imaging conditions not available on standard mammographic units.