Monte Carlo MCNP Research Articles

Neutron background signal in superheated droplet detectors of the Phase II SIMPLE dark matter search

The simulation of the neutron background for Phase II of the SIMPLE direct dark matter search experiment is fully reported with various improvements relative to previous estimates. The model employs the Monte Carlo MCNP neutron transport code, using as input a realistic geometry description, measured radioassays and material compositions, and tabulated (α,n) yields and spectra. Developments include the accounting of recoil energy distributions, consideration of additional reactions and materials and examination of the relevant (α,n) data. A thorough analysis of the simulation results is performed that addresses an increased number of non-statistical uncertainties. The referred omissions are seen to provide a net increase of 13% in the previously-reported background estimates whereas the non-statistical uncertainty rises to 25%. The final estimated recoil event rate is 0.372 ± 0.002 (stat.) ± 0.097 (non-stat.) evt/kgd resulting in insignificant changes over the results of the experiment.

The Lead-Based VENUS-F Facility: Status of the FREYA Project

The GUINEVERE project in the 6th European Framework Program (FP6) (1) aimed to check the methods for sub-criticality monitoring. To execute the project, the water- moderated thermal VENUS facility was modified into the lead fast VENUS-F facility in the period 2007-2010. To prove the reliability of the reactivity monitoring methods, first of all a critical reference configuration was assembled and characterized by measurements of criticality, power distribution, and spectral indexes. These experiments were communicated for benchmarking at ISRD-14 (2). The Monte Carlo MCNP 5-1.60 code with the JEFF 3.1.2 data library is used to perform simulations of the VENUS-F core, in particular to obtain Calculated-to-Experimental ratios (C/E) for fission rates and spectral indices. A sensitivity study is performed focusing on the impact of global and local parameters on C/E. In most cases C/E is close to unity within the uncertainties. Only a few exceptions were found, e.g. for the F28/F25 spectral index (3). In order to investigate the discrepancies, a new measurement campaign with the same critical configuration was included in the currently ongoing FREYA project in FP7 (4). The facility status, experimental plans, and the sensitivity study are presented in this paper.

Monte Carlo based dosimetry for neutron capture therapy of brain tumors

Boron Neutron Capture Therapy (BNCT) is a biologically targeted, radiation therapy for cancer which combines neutron irradiation with a tumor targeting agent labeled with a boron10 having a high thermal neutron capture cross section. The tumor area is subjected to the neutron irradiation. After a thermal neutron capture, the excited 11 B nucleus fissions into an alpha particle and lithium recoil nucleus. The high Linear Energy Transfer (LET) emitted particles deposit their energy in a range of about 10μm, which is of the same order of cell diameter [1], at the same time other reactions due to neutron activation with body component are produced. In-phantom measurement of physical dose distribution is very important for BNCT planning validation. Determination of total absorbed dose requires complex calculations which were carried out using the Monte Carlo MCNP code [2].

Monte Carlo Simulation Study of a Differential Calorimeter Measuring the Nuclear Heating in Material Testing Reactors

The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.

Dose Calculation for Radiation Safety Assessment of Plasma Focus 2.5 kJ

A study was conducted to dose analysis of human body organs of ORNL mathematical phantom as a personnel in radiation field of the 2.5 kJ plasma focus device. The dose calculations were performed for 108 fusion neutrons per shot (with energy of 2.45 MeV) from deuterium and deuterium fusion (D–D) using Monte Carlo MCNP code. Based on the results, assuming 2600 shots/year (10 shots/day), equivalent and effective dose obtained by operator in different distances from plasma focus device is less than the determined yearly working dose which is assumed 20 mSv/year. The calculations showed that for working in close distances to PF (distance about 50 cm) or in the PF radiation field with more than 10 shots/day, Polyethylene (with 5 % 10B) with 10 cm in thickness is sufficient to reduce effective dose less than 20 mSv/year.

Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200) that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models) Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code. We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon extended approach on sub-group method. The results of K inf , in case of fresh seed fuel pin which is used in thorium/uranium PWR, show that a high accuracy is obtained by using some specific self-shielding modules. It is also shown that for the implemented self-shielding models DRAGON library is more reliable than WIMS-D5 library, and that applying Livolant-Jeanpierre Normalization scheme is essential with the equivalence and dilution self-shielding method.

EDEL: ENEA DOSEMETER FOR EYE LENS.

Since the publication of International Commission on Radiological Protection statement in 2011 on tissue reaction, eye lens radiation protection played an important role in exposed personnel dosimetry. For this reason, the Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) Individual Monitoring Service decided to study a prototype to fulfil specific requests (e.g. for survey in interventional department and intercomparisons). On the basis of such preliminary investigation, a new eye lens dosemeter was developed. The new dosemeter, named EDEL (ENEA Dosemeter for Eye Lens), was characterised in terms of Hp(3), the operational quantity related to eye lens monitoring. The investigation was performed experimentally and optimised using the Monte Carlo MCNP6 code. The new prototype was thought to fulfil two main requests: the reliability of the dosimetric data and the portability of the dosemeter itself. The new dosemeter will soon be supplied to the collaborating hospitals for workplace test measurements.

VVER-1000 cross-section library generation for ORIGEN-II based on MCNP calculations

Most nuclear burnup, depletion and radioactive decay computational codes, such as ORIGEN-II, are geometry independent and so called point codes. Consequently, the effect of reactor geometry and other reactor conditions are compiled in their own cross-section library. For several types of reactor such as PWR or BWR, the cross-section libraries are provided for ORIGEN-II code, but for Bushehr nuclear reactor which is a VVER-1000 type reactor, such cross-section library does not exist in the code. In this study, attempts have been made to generate a cross-section library for VVER-1000 type reactor. In this study, attempts have been made to generate a cross-section library for VVER-1000 type reactor. In order to model the Bushehr Nuclear reactor (BNPP) core and generate a cross-section library suitable for ORIGEN-II calculations, MCNP Monte-Carlo code was used. The validation tests were carried out by comparing the burnup results of the new library with the burnup results obtained from MCNP calculations and burnup, depletion and radioactive decay reported in BNPP reference data.

Characterization of a high-purity germanium (HPGe) detector through Monte Carlo simulation and nonlinear least squares estimation

In this paper, we describe a method for determining the full-energy peak efficiency of a high-purity germanium detector using Monte Carlo MCNP simulation with the detector model refined through nonlinear least-squares optimization using experimental measurements. The approach allows for an accurate determination of efficiency curves in the absence of an accurate characterization of the detector by the manufacturer, usually a prerequisite for reliable Monte Carlo simulations.

Studying the effect of the lung size variation on dosimetry and systematic parameters of FUM-IVNAA facility

The aim of this paper was quantifying the uncertainties of dosimetry parameters and systematic factors of an in vivo neutron activation analysis (IVNAA) facility for neutron-irradiated individuals with different lung sizes. A series of statistical phantoms with adjustable lungs were modeled in IVNAA facility using Monte Carlo MCNP code. Obtained results showed that the fluctuation of the organ-absorbed dose and effective dose due to the lung size variation is negligible (less than 3 %). Also, changing the lung size will not significantly affect the systematic parameters of the facility.

Neutronic calculations for CANDU thorium systems using Monte Carlo techniques

In this paper, we have investigated the prospects of exploiting the rich world thorium reserves using Canada Deuterium Uranium (CANDU) reactors. The analysis is performed using the Monte Carlo MCNP code in order to understand how much time the reactor is in criticality conduction. Four different fuel compositions have been selected for analysis. We have obtained the infinite multiplication factor, k∞, under full power operation of the reactor over 8 years. The neutronic flux distribution in the full core reactor has already been investigated.

Unique formulations in TITAN and PENTRAN for medical physics applications

Special algorithms for application to medical physical problems have been developed and implemented into the TITAN and PENTRAN 3-D parallel transport codes. TITAN is a 3-D parallel transport code with hybrid algorithms including 1) Sn and Characteristics, and 2) Sn with a fictitious quadrature set and ray-tracing. PENTRAN is a 3-D parallel transport code with adaptive differencing Sn formulation, full domain decomposition, and different angular quadrature types (including a characteristic option) with ordinate splitting. In this paper, we will discuss new algorithms developed in TITAN for image reconstruction in Single Photon Emission Computed Tomography (SPECT). It is demonstrated that TITAN projection images are in excellent agreement with the Monte Carlo MCNP5 predictions, while its computation time was up to 2787 times faster than the MCNP5 computation time for a 0.98° collimator acceptance angle. We will also discuss a novel algorithm developed in PENTRAN for dose calculations in heterogeneous, voxelized phantoms or other geometries. The new algorithm called EDK-Sn, or “Electron Dose Kernel-Discrete Ordinates (EDK-Sn)” yielded a speedup of ~8 over traditional highly parallel Monte Carlo calculation times, with <7% difference in dose among different organs (or smaller given stochastic uncertainties).

Dosimetric Evaluation of Heterogeneities in Small Circular Fields of 6 MV Photon Beams with EBT2 and EDR2 Films: Comparison with Monte Carlo Calculation

The influence of dose variation due to heterogeneities in narrow photon beams used in stereotactic radiosurgery has been investigated. Since the lateral electronic disequilibrium and existence of steep dose gradients in small fields and the presence of heterogeneities can intensify these problems, in this study the effects of heterogeneities on 6 MV small photon beams produced by circular cone collimators with 5, 10, 15, 20 and 30 mm diameters are investigated. The heterogeneities include 3 cm Cork with density of 0.2 g/cm3 instead of lung and 3 cm Polytetrafluoroethylene (P.T.F.E) with density of 2.2 g/cm3 as bone. The measurements were carried out with EBT2 gafchromic and EDR2 radiographic films. Simulation was done by MCNP Monte Carlo Code (MCNP5). The depth dose curves in heterogeneous phantom were compared with homogeneous phantom. A good agreement was obtained within film and Monte Carlo calculations in presence of low density heterogeneity and also in the presence of high density heterogeneity. Monte Carlo results showed good agreement after stopping power correction.

Concept design of an illicit material detection system

This paper covers the concept design of an illicit material detection system under the project: “Aviation security-Design of a Nuclear Instrument for Contraband Detection” established by King Abdulaziz University. The detection system is composed of two units. The first unit is an X-ray machine which will indicate a location of interest and the second unit is a neutron based system. The motivation of this project is that the design is based on finding the location of the suspicious object inside the luggage by the first unit. This location will be the point of interest for the neutron based detection system. Since checks by the neutron based system are carried out for only suspicious baggage inspected by X-ray inspection system, occasional delays in the neutron based detection system should be acceptable. In this paper, the engineering description of the system is descried and the neutron based system is simulated using Monte Carlo MCNP code.

Thermoluminescent and Monte Carlo dosimetry of a new 170Tm brachytherapy source

In this Study characteristics of a new 170Tm brachytherapy seed using thermoluminescent dosimeter and also the Monte Carlo simulations to evaluate between calculated and measured values was determined. Titanium tube contained Tm(NO3)3 powders bombardment at the Tehran Research Reactor (TRR) for a period of 7 days at a flux of 2–3 × 1013 neutrons/cm2 s. To obtain the radial dose function, g(r), and the anisotropy function, F(r, θ), according to the AAPM TG-43U1 recommendations, 30 cm × 30 cm × 15 cm phantoms of Perspex slabs were used. Brachytherapy dose distributions were simulated with the MCNP5 Monte Carlo (MC) radiation transport code. The MCPLIB04 photon cross-section library was applied using data from ENDF/B-VI. Cell-heating tally, F6 was employed to calculate absorbed dose in two separate runs for both beta and gamma particles. The calculated dose rate constant for the HDR source was found to be 1.113 ± 0.021 cGyU−1 h−1. Nominal uncertainty in the measured and calculated radial dose functions, g(r), for the IR-170Tm source in Perspex is tabulated is approximately 6% (ranging from 2% to 9%). The anisotropy function, F(r, θ), of the IR-170Tm source was measured at radial distances of r = 1.5, 2, 3, 5 cm relative to the seed center, and polar angles θ ranging from 0° to 330° in 30° increments.

Photoneutron contamination from an 18 MV Saturne medical linear accelerator in the treatment room

Dose escalation with high-energy X rays of medical linear accelerators (linacs) in radiotherapy offers several distinct advantages over the lower energy photons. However, owing to photoneutron reactions, interaction of high-energy photons (>8 MV) with various high-Z nuclei of the materials in the linac head components produces unavoidable neutrons. The aim of this study was to evaluate the photoneutron dose equivalent per unit therapeutic X-ray dose of 18 MV, GE Saturne 20 linac in the treatment room using Monte Carlo (MC) MCNP linac head full simulation as well as thermoluminescence dosemeter measurements. This machine is one of the old linac models manufactured by General Electric Company; however, it is widely used in the developing countries because of low cost and simple maintenance for radiotherapy applications. The results showed a significant photoneutron dose from Saturne 20 linac head components especially at distances near the linac head (<150 cm). Results of this work could be used in several applications, especially designing bunker and entrance door shielding against neutrons produced by photoneutron reactions in GE Saturne 20. However, a detailed cost optimisation for a specific room would require a dedicated calculation.

Monte Carlo investigation of the effect of small cutouts on beam profile parameters of 12 and 14 MeV electron beams

Cutouts, which are used as field-shaping shield, affect several electron beam parameters. These effects are more observable for small field sizes and high energy electron beams. Owing to the fact that small fields prevent the lateral scatter equilibrium, at higher energies larger field radius is required for the establishment of lateral equilibrium. The profile curves are derived from circular, triangular, and square cutout shapes and size placed in a 10 × 10 cm 2 electron applicator. These profile curves are obtained using parallel plane type ion chamber at the R 100 , R 90 , R 80 and R 50 depths. Correspondingly, the source surface distance is 100 cm. In this study MCNP Monte Carlo (MC) simulation was used to compare Percentage Depth Dose (PDD) and Profile of electron beams. Monte Carlo and measured results showed a good compliance for PDD and beam profile. The measurements and calculations showed that as the field width decreases, the Flatness and Penumbra Ratio also decreases. In other words, flatter plateau was available for larger fields. Also the Coverage Ratio for each of the profiles is presented. The flatness and symmetry values for triangle shapes were greater than the two other shapes. Knowledge of these changes are significant in radiation therapy. Accordingly, a comparison between the Monte Carlo data and the measured results can be beneficial for treatment simulation and development of treatment planning systems. ► Mesh Tally 1 and pedep keyword were used to calculate the PDD and profile values. ► In measurement the coverage for larger fields and fewer doses are better. ► By increasing the depth, the flatness and symmetry values were increased. ► The worst flatness and symmetry (between 3 compared shapes) belonged to triangle. ► The given Penumbra and Coverage Ratio can be helpful for PTV margin and coverage.

Dosimetry of MammoSite® applicator: Comparison between Monte Carlo simulation, measurements, and treatment planning calculation

To investigate the dosimetric characteristics of accelerated partial breast irradiation technique by MammoSite® applicator using thermoluminescent dosimeter (TLD) and Monte Carlo simulation to comparing them with treatment planning system calculation for planning target volume (PTV) and organs at risk such as skin, lung and chest wall. The Monte Carlo MCNP-5 code was used to simulate dose rate in the PTV that is a MammoSite® balloon with 1 cm margin around it. Experimental dosimetry was carried out within a female-equivalent chest phantom with TLD dosimeter after insertion of 192 Ir source into the MammoSite® applicator. Three dimensional planning (TP) was done for dose delivery to the specific points within the phantom by means of FlexiPlan software. Statistical comparisons were done between TP calculation, Monte Carlo simulation and TLD. Our results showed good agreement for surface doses between simulation and measurement. The mean skin dose for the simulation and TLD result was 61.7% and 56.8% of prescription dose, respectively. The maximum dose to the chest wall for Monte Carlo and TLD were 114.4% and 111.8% of prescription dose, respectively. The maximum dose to the lung for Monte Carlo and TLD results were 28.4% and 27.3% of prescription dose, respectively. Using Monte Carlo simulation and an average female chest phantom, it was possible to demonstrate the accuracy on the calculated dose rate in the PTV of a MammoSite® dose delivery system with 192 Ir HDR sources. The results showed acceptable agreement between simulation, treatment planning, and experimental dosimetry results.

Monte Carlo MCNP modeling of a HPGe detector and its efficiency for extended sample geometry

The activity of an extended 241Am sample with complex shielding and geometry has been measured by an 80% HPGe detector, after its irradiation by a neutron beam at 10.4 MeV, in order to determine the cross section of the reaction 241Am(n,2n)240Am. Due to the complexity of sample’s geometry, the estimation of the detector’s efficiency has been achieved by MCNP5 Monte Carlo simulations. The simulations have been gradually evolved in order to define the detector model, the experimental setup geometry (geometry definition), the sample’s density and finally the detector’s efficiency for the irradiated 241Am sample.

New technique for effective dose estimation using Monte Carlo simulation for the patients undergoing radioiodine therapy

We presume that radioiodine capsule, used for therapy, stays in stomach for 15min before the absorption starts when large amount of radioactivity needlessly expose a part of stomach and several surrounding organs. Investigations were performed at Institute of Pathophysiology and Nuclear Medicine, Faculty of Medicine, Ss. Cyril and Methodius University, Skopje, Macedonia. Eighty-seven patients were reviewed, age between 21 and 73. Patients were divided in 20 groups receiving average dose per group from 1813MBq to 6105MBq. We used Monte Carlo MCNP 4b code and gave the results only for administrated activity of 6105MBq. Calculated results are given for imparted energy per transformation, dose equivalent and effective doses for different organs. The effective doses were between 9.17nSv (for bone surface) to 122.4mSv (for stomach). As the total dose is estimated to be 126.73mSv it is obvious that the highest part is received by stomach.