Carlo N-Particle Transport Code Version Research Articles

New operational quantities (RBE × D p lens/Φ) for eye lens neutron dosimetry as a function of energy and angle of incidence in the ICRU 95 formalism

This paper is a continuation of a study published recently by the authors. It presents and discusses computed personal absorbed dose in the lens of the eye (D p lens/Φ), and a relative biological effectiveness (RBE)-weighted absorbed dose (in terms of an newly proposed operational quantity RBE × D p lens/Φ), conversion coefficients for the lens of the eye for neutron exposure at incident energies from thermal to ∼20 MeV and at angles of incidence from 0° to 90° in 15° increments, at 180° and for rotational incidence irradiation geometry (from 0° to 360° in 5° increments). These conversion coefficients were obtained from a simulation model developed for this study that contains the stylised eye model, embedded in the adult UF-ORNL mathematical phantom, whereby the previously stated RBE-weighted absorbed dose was obtained using the proposed RBE versus neutron energy distribution compiled in a previous paper by the authors. The simulations carried out for this study using the Monte Carlo N-Particle transport code version 6.2, were conducted in a realistic human eye model, for the left and right sensitive and whole volume of the lens of the eye, considering the recent proposed redefinition of the operational quantities for external radiation exposure in International Commission on Radiation Units and Measurements (ICRU) report 95. A comprehensive set of tabulated data for neutron fluence-to-dose conversion coefficients (D p lens/Φ in pGy cm2) and RBE-weighted absorbed dose (RBE × D p lens/Φ in pGy cm2) conversion coefficients is included in this paper as a function of incident neutron energy and angle of incidence. Data for D p lens/Φ (pGy cm2) are compared to similar data from the literature for validation of our model. Data for RBE × D p lens/Φ (in pGy cm2), were also compared to the equivalent operational quantity H p(3,α)/Φ (in pSv cm2) conversion coefficients calculated at 3 mm depth in a cylindrical phantom for different incident neutron energies and angles of incidence from 0° to 75° in 15° increments to demonstrate the relevance of this newly proposed operational quantity for doses resulting in tissue reactions (deterministic effects) which should be quoted in Gray (RBE-weighted absorbed dose, RBE × D (Gy)), rather than Sievert (Sv) which is reserved for stochastic effects. The current neutron weighted absorbed dose (RBE × D p lens) is proposed for the tissue reactions in the eye-lens for neutron radiation as per National Council on Radiation Protection and Measurements report 180 and in line with the recent proposal for the review and revision of the System of Radiological Protection to Keeping the International Commission on Radiological Protection (ICRP) recommendations fit for purpose. This method would bring better alignment between the dose limits in ICRP 118 and the new operational quantity consistent with the units of the new eye-lens dose limits without being overly conservative. The utilization of the proposed new operational quantities, as outlined in ICRU 95, has the potential to address the ongoing challenge in enforcing regulatory limits for neutron eye dose, specifically the use of Gy instead of Sv. It should be noted that the applicability of this will vary from country to country as in many countries the legislation is likely to mandate the use of H p(3) until the regulation is amended. This approach can serve as an interim solution while awaiting the issuance of the new ICRP general recommendations, which is expected to take several years. Implementing the new operational quantities can contribute to enhancing the accuracy and effectiveness of neutron eye dose limit enforcement.

Radiation characterizations of two isotopic neutron sources merging in one irradiator for experimental applications in the laboratory

Neutron sources are utilized for different aims, such as studying the material’s internal structure, investigating the materials’ crystal structures, and neutron therapy. Using more than one isotopic neutron source in some aims could be beneficial. Herein, a neutron irradiator consisting of two 241Am-9Be neutron sources has been designed. Radiation characterizations, including the neutron fluence and dose rate, were investigated. The Monte Carlo simulation code Monte Carlo N-particle transport code version 5 (MCNP5) was used to conduct all the required investigations to achieve an optimal design. Two factors were considered; the first was the dose rate due to both neutron and gamma radiations to meet radiation protection requirements; MCNP5 was used to determine the adequate thickness of the shielding material. The second parameter is the neutron fluence. Two irradiation sites have been proposed, one for fast neutron irradiation and the other for thermal neutron irradiation. By utilizing MCNP5, we have determined the radiation characterization at the two sites to be used as the irradiation sites. After the optimal design had been produced, the thermal and the epithermal neutron fluences inside the two irradiation sites were determined experimentally. The experimental results showed a perfect agreement with the simulation ones. The neutron irradiator designed from merged two isotopic neutron sources could be a benefit in the neutron radiation applications in the labs.

Evaluation of Radiological Effects on the Aptamers to Remove Ionic Radionuclides in the Liquid Radioactive Waste

Background: Aptamers are currently being used in various fields including medical treatments due to their characteristics of selectively binding to specific molecules. Due to their special characteristics, the aptamers are expected to be used to remove radionuclides from a large amount of liquid radioactive waste generated during the decommissioning of nuclear power plants. The radiological effects on the aptamers should be evaluated to ensure their integrity for the application of a radionuclide removal technique.Materials and Methods: In this study, Monte Carlo N-Particle transport code version 6 (MCNP6) and Monte Carlo damage simulation (MCDS) codes were employed to evaluate the radiological effects on the aptamers. MCNP6 was used to evaluate the secondary electron spectrum and the absorbed dose in a medium. MCDS was used to calculate the DNA damage by using the secondary electron spectrum and the absorbed dose. Binding experiments were conducted to indirectly verify the results derived by MCNP6 and MCDS calculations.Results and Discussion: Damage yields of about 5.00×10<sup>-4</sup> were calculated for 100 bp aptamer due to the radiation dose of 1 Gy. In experiments with radioactive materials, the results that the removal rate of the radioactive <sup>60</sup>Co by the aptamer is the same with the non-radioactive <sup>59</sup>Co prove the accuracy of the previous DNA damage calculation.Conclusion: The evaluation results suggest that only very small fraction of significant number of the aptamers will be damaged by the radioactive materials in the liquid radioactive waste. Keywords: Aptamer, Liquid Radioactive Waste, Decommissioning, Decontamination, Radiological Damage

Physical, structural, elastic and optical investigations on Dy3+ ions doped boro-tellurite glasses for radiation attenuation application

A set of Dy3+ ions doped boro-tellurite glasses is prepared using the melt quenching method and its radiation resisting aptitude is estimated through various structural and elastic features. The results were interpreted due to the effect of TeO2 concentration in the glass systems. The observed molar volume (Vm) values reduced with the rise in the tellurium content with the creation of more bridging oxygens in the structure. The cross-link density rises the rigidity of the glass and its density. Additionally, boron-boron separation and tellurium-tellurium separation, and optical band gap were computed to prove the connectivity of the system which is noted to reduce with the rise in the tellurium content. Glass with 40% TeO2 is observed to hold high elastic moduli and consequently high density and connectivity which are vital requirements for effective radiation shielding. Moreover, the Monte Carlo N-Particle transport code version 5 is used to qualify the γ-ray shielding properties. The qualification of the shielding properties showed that the linear attenuation coefficient was enhanced by a factor of 80% from 0.248 cm−1 to 0.358 cm−1 by raising the TeO2 concentration between 0 and 40 wt %, respectively. The enhancement on linear attenuation coefficient was reflected on the half value thickness (Δ0.5, cm) and thickness equivalent (Δeq, cm) of the fabricated samples where Δ0.5 and Δeq values reduced with raising the TeO2 concentration between 0 and 40 wt %.

Estimation of optimized timely system matrix with improved image quality in iterative reconstruction algorithm: A simulation study

The system matrix (SM) being a main part of statistical image reconstruction algorithms establishes relationship between the object and projection space. The aim was to determine it in a short duration time, towards obtaining the best quality of contrast images. In this study, a new analytical method based on Cavalieri's principle as subdividing common regions has been proposed in which the precision of the amounts of estimated areas was improved by increasing the number of divisions (NOD), and consequently the total SM's time was increased. An important issue is the tradeoff between the NODs and computational time. For this purpose, a Monte Carlo simulated Jaszczak phantom study was performed by the Monte Carlo N-Particle transport code version 5 (MCNP5) in which the tomographic images of resolution and contrast phantoms were reconstructed by maximum likelihood expectation maximization (MLEM) algorithm, and the influence of NODs variations was investigated. The results show that the lowest and best quality have been obtained at the NODs of 0 and 8, respectively and in the optimum case, the SM's total time at NOD of 8 was 925 s, which was much lower than those of the conventional Monte Carlo simulations and experimental test.

Feasibility study of artifacts on the neutron logging curve near the boundary of layers

The neutron borehole probe equipped with a neutron source and a system of neutron detectors is commonly used to determine porosity of geological formations. Transport of neutrons through media relies on their interaction with atomic nuclei, mainly with the process of scattering and absorption. Curves recorded by detectors for a given geological layer bordered by a medium with other neutron properties form the so-called anomalies. The course of the anomaly mainly depends on neutron properties of the bordering media and also on the measurement geometry, borehole properties, etc. The paper focuses mainly on an influence of neutron properties of bordered rock media on the course of the anomaly curve, especially near the media boundaries. In some cases of bordering media, characterized by specific neutron properties, the course of anomalies can be significantly disturbed, forming specific artifacts. The paper explains that the presence of artifacts is closely related to the properties of bordered media for both fast and thermal neutrons, i.e., the slowing down length of fast neutrons and the absorption cross section of thermal neutrons. Artifacts can be the cause of misinterpretation of the neutron rock porosity probe measurement. All considerations presented in this paper are based on results obtained from MC calculations performed using MCNP5 code (MCNP Team in X-5 Monte Carlo Team, MCNP—a general Monte Carlo N-particle transport code version 5. Los Alamos National Laboratory LA-UR-03-1987, 2008).

Nuclear analysis of the Single Module Segment WCLL DEMO

The latest design of the Water Cooled Lithium Lead blanket for the European DEMO reactor is based on a Single Module Segment (SMS) concept for the blanket segments: the removal of the gaps between the modules (present in the former Multi Module structure design), with the consequent reduction of the neutron streaming, improves the tritium production and the shielding capabilities. A nuclear analysis has been carried out with the Monte Carlo N-Particle transport code version 5 (MCNP5) and the Joint Evaluated Fission and Fusion nuclear data libraries version 3.2 (JEFF 3.2). A detailed three-dimensional MCNP DEMO model with the Single Module layout has been generated. Three-dimensional neutron and gamma transport simulations have been carried out in order to assess the tritium production, nuclear power deposition in each subcomponent of the blanket and nuclear heating density, neutron flux, neutron damage (dpa) and Helium production, relevant for the design of the system. The results confirm the fulfilment of tritium self-sufficiency and shielding requirements of DEMO. Furthermore, a parametric analysis has been carried out by varying the thickness of the first wall tungsten layer to evaluate how the blanket performances are influenced by the tungsten layer thickness.

Comparison of thermoluminescence energy response of optical fiber subject to photon irradiation between Monte Carlo simulation and experiments

This study is focused on energy response of Ge-doped silicon dioxide optical fiber subjected to photon irradiation. The TL responses for photon energies, ranging from 20 keV to 20MeV, were investigated as energy absorbed in the TL materials of Ge-doped silicon dioxide optical fiber. The simulation was performed using Monte Carlo N-particle transport code version 5 (MCNP5). The input parameters included in this study are source information, geometry specification, material information and tallies F6. Comparisons of energy response were done between simulation and previous results of experiments. A flat response can only be seen in the energy range of 200 keV to 10 MeV.

Evaluation of internal contamination levels after a radiological dispersal device incident using portal monitors

Following a radioactive dispersal device (RDD) incident, it may be necessary to evaluate the internal contamination levels of a large number of potentially affected individuals to determine if immediate medical follow-up is necessary. Since the current laboratory capacity to screen for internal contamination is limited, rapid field screening methods can be useful in prioritising individuals. This study evaluated the suitability of a radiation portal monitor for such screening. A model of the portal monitor was created for use with models of six anthropomorphic phantoms in Monte Carlo N-Particle Transport Code Version 5 (MCNP) X-5 Monte Carlo Team (MCNP-A General Monte Carlo N-Particle Transport Code Version 5. LA-CP-03-0245. Vol. 2. Los Alamos National Laboratory, 2004.). The count rates of the portal monitor were simulated for inhalation and ingestion of likely radionuclides from an RDD for each of the phantoms. The time-dependant organ concentrations of the radionuclides were determined using Dose and Risk Calculation Software Eckerman, Leggett, Cristy, Nelson, Ryman, Sjoreen and Ward (Dose and Risk Calculation Software Ver. 8.4. ORNL/TM-2001/190. Oak Ridge National Laboratory, 2006.). Portal monitor count rates corresponding to a committed effective dose E(50) of 10 mSv are reported.

Thermoluminescence energy response of TLD-100 subjected to photon irradiation using Monte Carlo N-particle transport code version 5

Useful TL properties of TLD-100 that is an excellent candidate for using in TL dosimetry of ionizing radiation are demonstrated. This study is focused on response of TLD-100 subjected to photon irradiation. The thermoluminescence (TL) response of TLD-100 subject to various photon energy, ranging from 20 keV to 6 MeV, was investigated as energy absorbed in the TL material using Monte Carlo N-Particle transport code version 5 (MCNP5). The input parameters included in this study are experimental geometry specification, source information, material information, and tallies. Tally F6 is used in this simulation. The results from MCNP5 simulation show good agreement with previous experimental data. However, the data obtained from the simulation are greater than the experimental data especially in lower energy ranges.

Proof and implementation of the stochastic formula for ideal gas, energy dependent scattering kernel

The ideal gas, scattering kernel for heavy nuclei with pronounced resonances was developed [Rothenstein, W., Dagan, R., 1998. Ann. Nucl. Energy 25, 209–222], proved and implemented [Rothenstein, W., 2004 Ann. Nucl. Energy 31, 9–23] in the data processing code NJOY [Macfarlane, R.E., Muir, D.W., 1994. The NJOY Nuclear Data Processing System Version 91, LA-12740-M] from which the scattering probability tables were prepared [Dagan, R., 2005. Ann. Nucl. Energy 32, 367–377]. Those tables were introduced to the well known MCNP code [X-5 Monte Carlo Team. MCNP – A General Monte Carlo N-Particle Transport Code version 5 LA-UR-03-1987 code] via the “mt” input cards in the same manner as it is done for light nuclei in the thermal energy range. In this study we present an alternative methodology for solving the double differential energy dependent scattering kernel which is based solely on stochastic consideration as far as the scattering probabilities are concerned. The solution scheme is based on an alternative rejection scheme suggested by Rothenstein [Rothenstein, W. ENS conference 1994 Tel Aviv]. Based on comparison with the above mentioned analytical (probability S( α, β)-tables) approach it is confirmed that the suggested rejection scheme provides accurate results. The uncertainty concerning the magnitude of the bias due to the enhanced multiple rejections during the sampling procedure are proved to lie within 1–2 standard deviations for all practical cases that were analysed.

Analysis of x-ray spectrum obtained in electron cyclotron resonance x-ray source

The analysis of the x-ray spectrum obtained in electron cyclotron resonance (ECR) x-ray source is carried out. Assuming single-particle motion, the electron acceleration and its final energy are calculated for TE111 cylindrical cavity field and uniform external dc magnetic field. In the calculation, initial coordinates of 40 000 electrons were uniformly selected over the central plane of the cavity using random number generator. The final energy of each electron when it hits the wall is stored and the electron energy distribution is obtained. Using the general purpose Monte Carlo N-particle transport code version 4A, the geometry of the ECR x-ray source is modeled. The x-ray energy spectrum is calculated for the geometry model and the numerically calculated electron energy distribution. The calculated x-ray spectrum is compared with the experimentally measured x-ray spectrum.

Ford motor company NDE facility shielding design

Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations.

Assessment of90Sr Concentration in Dental Tissue using Thin-Layer Beta-Particle Detectors and Verification with Numerical Calculations

Electron paramagnetic resonance (EPR) measurements of tooth enamel can be used as an individual biological dosimeter for external dose assessment. However, the presence of 90Sr in the tooth tissues makes the task of interpreting EPR tooth dosimetry more complicated. The determination of the dose contribution of incorporated 90Sr in calcified tissue to the total dose measured by EPR is one of the main aspects of correct interpretation of EPR tooth dosimetry. In this work, experimental and numerical calculations were performed to convert the measured beta-particle dose rate to 90Sr concentration in calcified tissue. The cumulative beta-particle dose was measured by exposing artificially contaminated dentin and enamel to thin-layer alpha-Al2O3:C detectors in two different exposure geometries. Numerical calculations were performed for experimental exposure conditions using calculations of electron transport and secondary photons [Monte Carlo n-Particle Transport code version 4C2 (MCNP)]. Numerical calculations were performed to optimize the sample size and exposure geometry. The applicability of two different exposure conditions to be used in routine analysis was tested. Comparison of the computational and experimental results demonstrated very good agreement.

Neutron dose equivalent estimation from the specific activity of [formula omitted

The exposed neutron dose from the JCO criticality accident has been preliminarily estimated from data on the specific activity of induced 51 Cr using the Monte Carlo N-Particle Transport Code version 4B. The dependence on ground distance of the measured specific activities is fitted to the results of the transport calculation. The neutron dose equivalents for ground distances of 130–450 m from the JCO range from 36 to 1 mSv.