Monte Carlo Code MCNP Research Articles

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.

Annular fuel neutronic analysis

The annular fuel was proposed to increase the power density while keeping or improving the safety margins, given the considerably lower surface heat flux and lower fuel temperature than conventional solid fuel. In this paper, neutronic characteristics of this new fuel design were analysed using the new transport code of the Westinghouse, called PARAGON2, using Ultra-Fine Energy Mesh Library for multi-group cross section calculations. The bibliographical review showed there is a lack of deterministic transport codes capable to model such fuel with acceptable accuracy, because the equivalence relations for heterogeneous resonance integrals adopted in the standard codes were formulated only for solid cylindrical rods. Therefore, the main objective of this paper is to perform neutronic analysis for annular fuel using PARAGON2, evaluating the use of Ultra-Fine Energy Mesh Library to eliminate resonance self-shielding calculation. The results of this paper analysed the power distributions, the ratio of U-238 capture rate to U-235 fission rate and behaviour of the infinite multiplication factor for annular fuel and compared with the Monte Carlo code MCNP.

Neutronics Assessment of the Spatial Distributions of the Nuclear Loads on the DEMO Divertor ITER-like Targets: Comparison between the WCLL and HCPB Blanket

The Plasma Facing Components (PFCs) of the divertor target contribute to the fundamental functions of heat removal and particle exhaust during fusion operation, being subjected to a very hostile and complex loading environment characterized by intense particles bombardment, high heat fluxes (HHF), varying stresses loads and a significant neutron irradiation. The development of a well-designed divertor target, which represents a crucial step in the realization of DEMO, needs the assessment of all these loads as accurately as possible, to provide pivotal data and indications for the design and structural performance prediction of the PFCs. In a particular way, this study is fully devoted to the comprehension of the distributions on the divertor target of the main nuclear loads due to neutron irradiation, performed for the first time using an extremely detailed approach. This work has been carried-out considering the latest configuration of the DEMO reactor, including the updated design of the divertor and ITER-Like PFCs geometry, varying the blanket layout (Water Cooled Lithium Lead—WCLL and Helium Cooled Pebble Bed—HCPB), thus evaluating the impact of the different blanket concept on the above-mentioned distributions. Neutronics analyses have been performed with MCNP5 Monte Carlo code and JEFF3.3 nuclear data libraries. 3D DEMO MCNP models have been created, focusing in particular on a thorough representation of the divertor and PFCs, allowing for the assessment of the distributions of the main nuclear loads: radiation damage (dpa/FPY), He-production rate (appm/FPY) and nuclear heating density (W/cm3) and total nuclear power deposition (MW). These results are presented by means of 2D maps and plots for each PFCs sub-component both for WCLL and HCPB blanket case: W-monoblocks, Cu-interlayers\CuCrZr-pipe and PFC-CB (Cassette Body) supports made of Eurofer steel.

Benchmarking of stray neutron fields produced by synchrocyclotrons and synchrotrons in compact protontherapy centers (CPTC) using MCNP6 Monte Carlo code

Proton therapy is an external radiotherapy using proton beams with energies between 70 and 230 MeV to treat some type of tumours with outstanding benefits, due to its energy transfer plot. There is a growing demand of facilities taking up small spaces and Compact Proton Therapy Centers (CPTC), with one or two treatment rooms, supposing the technical response of manufacturers to this request. A large amount of stray radiation is yielded in the interaction of proton beam used in therapy, neutrons mainly, hence, optimal design of shielding and verifications must be carried out in commissioning phases. Currently, almost 50 proton centers are under construction and start up in several countries, including ten in Spain.In the present work the effectiveness of shielding in two CPTC was verified with the Monte Carlo code MCNP6 by calculating the ambient dose equivalent, H*(10) due to secondary neutrons, outside the enclosures and walls of the center. The facilities modelled were the two centers currently operating in Spain, the first, since December 2019, with a superconductor synchrocyclotron, and the second, since March 2020, with a compact synchrotron. The geometry and materials are based on dimensions proposed a priori by the vendors, therefore, the paper is focused on check the suitability of the materials and thickness of the walls of the centers. Several models of the radiation sources were simulated, starting from a conservative assumptions, followed by more realistic scenarios. In all cases, the results reached for the ambient dose equivalent, H*(10), were below 1 mSv/year, which is the legal limit considered for the public in international references. Finally, considering that the recent ICRU Report 95 proposes changes in the operational quantities, the dose outside shieldingt has been evaluated in terms of the new next area surveillance quantity, H*, known as ambient dose, in the process of implementation.

Sensitivity analysis of gamma-ray shielding characteristics to the TiO2 concentration in the Bi2O3–ZnO–Pb3O4–Al2O3 glass sample based on the Monte Carlo method

In this paper, the MCNP Monte Carlo (MC) code, was adopted for evaluation the properties of protection against gamma-rays of the glass system with the composition of (55-x)Bi2O3–15Pb3O4–20Al2O3–10ZnO- xTiO2 with specific concentrations of (x = 0, 5, 10, 15, 20, 25, 30 and 35 mol percent). For this purpose, several parameters related to photon attenuation such as the, Half-Value Layer (HVL), Tenth-Value Layer (TVL), mean free range (MFP), mass attenuation coefficient (μm), effective atomic number (Zeff) and buildup factor (BF) was investigated at the different energy levels between 0.1 and 1.5 MeV. To verify the results of the simulation, the data obtained from the simulation were compared with the experimental data available in the XCOM database. The NIST-XCOM database and the results of the simulation in the present study were in good agreement. The percentage deviation (PD) between the XCOM database and the MCNP program for calculating the μm was less than 0.59% in most cases. The results show that compared to conventional protective materials such as concrete and lead, the new combination shows more effective attenuation parameters. The glass with the highest concentration of TiO2 has the most favorable state in terms of density compared to the investigated protective materials. The close agreement of μm measurements with the XCOM database of this research shows that MC modeling in this research is a valuable and valid method.

Nuclear Analysis of High-Power LIEBE Molten Target at CERN for the Production of Radioisotopes

To enhance the production of short-lived isotopes, higher beam powers are sought, which require targets able to accommodate them. One such target prototype is a liquid metal target LIEBE, developed at CERN. In this paper, a simulation of the proton beam interaction with the LIEBE target is presented. Simulations were performed by a series of proton transport calculations using the MCNP Monte Carlo code. The latest LIEBE target MCNP input was created in high-fidelity geometry, and the FENDL-3.1 cross-section data library was used. Flux and dose rate maps in the LIEBE target obtained from the simulations are presented in the paper. The maximum obtained dose around the target is roughly 361 Sv/h for gamma rays and 214 Sv/h for neutrons. The 70 MeV–100 µA proton beam penetrates roughly 7 mm deep into the liquid eutectic lead–bismuth. Based on this, further required changes to the LIEBE target can be evaluated.

In silico analysis of optimum photon energy spectra and beam parameters for iodine nanoparticle–aided orthovoltage radiation therapy of brain tumors

In this study, the effect of photon energy spectra on the dose enhancement factor (DEF) and other influencing parameters such as skull–tumor dose ratios were estimated for a confined tumor loaded with iodine nanoparticles (INPs). A mathematical brain phantom with a brain tumor was simulated by the MCNP6 Monte Carlo code. A validated commercial computed tomography model consisting of an X-ray tube, Al–Cu filters, and collimators was used to simulate the rotational conformal treatment of the tumor. INPs with a diameter of 20 nm and concentrations of 10–50 mg/g were introduced inside the tumor as homogenously distributed spherical particles. The dose distribution inside the phantom was scored for several orthovoltage beams with the peak voltages of 140, 200, and 320 kVp as well as two 3.5 and 6 MV megavoltage beams. A significant rise of DEF values with nanoparticle (NP) concentration for orthovoltage beams is revealed; no significant dose enhancement was obtained for megavoltage beams. The highest DEF and skull–tumor dose ratio were obtained for the 140 kVp beam which decreased with the number of directional fields. The clinically optimal plan for a brain tumor, with high DEFs of 2.81–2.24 and acceptable skull–tumor dose ratios of 0.61–0.51, would be feasible for treatment using 200 and 320 kVp beams, an iodine concentration of 20 mg/g, and 8–15 fields. Our calculations show that clinically significant radiation dose enhancements can be obtained for tumors loaded with INPs using orthovoltage beams. Optimal treatment regimens are feasible using a proper selection of photon beam spectrum and sufficient numbers of cross-firing beams. The limiting effect of skull bone could be minimized by increasing the number of radiation fields and the use of higher quality of orthovoltage beams.

ESTIMATION OF NEUTRON AND PROMPT PHOTON DOSE RATE DISTRIBUTION IN TMSR-500 USING MCNP6

Thorium Molten Salt Reactor-500 (TMSR-500), one of the Generation IV nuclear reactors, is designed by Thorcon International, Pte. Ltd, which is projected to be built in Indonesia. The reactor core is radially surrounded by B4C shielding, but not the upper part. As the silo hall sits above the reactor core and is accessible by reactor personnel, the dose rate must be calculated in the area to ensure the workers receive an annual dose below the acceptable limit. The dose rate from neutrons and photons as the result of fission reactions are the only sources to be calculated in this research, without taking the source from fission products into account. This research aims to obtain the dose rate distribution of neutrons and prompt photons using Monte Carlo code MCNP6. The reactor was assumed to operate at a nominal thermal power of 557 MWth. Dose rate calculation was obtained from flux Tally F4 and converted into dose rate using Dose Energy Dose Function (DEDF) factor. Conversion factors of flux to the dose were based on ICRP-21 and ANSI/ANS-6.1.1 1977. The result of the calculations showed that the distribution of neutron and prompt photon fluxes does not reach the silo hall.

Neutronic and Shielding Analyses for the DTT Electronics

The radiation effect on electronic devices is a critical issue for nuclear facilities, accelerators, and spatial missions as well as for fusion machines. Phenomena related to particle interactions have significant impact on the normal functioning of such devices: they could arise progressively during machine lifetime due to the cumulated ionization [total ionizing dose (TID)], cumulated atomic displacement (dpa) or it can appear instantaneously after a single highly ionizing particle interaction [single-event effect (SEE)]. The divertor tokamak test (DTT) facility is a tokamak machine, which will produce, in its high-performance phase, up to <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$1.5\times 1017$ </tex-math></inline-formula> n/s at 2.45 MeV from deuterium–deuterium reactions plus about 1% of 14.1-MeV neutrons produced by the triton burn-up inside the plasma. Such significant neutron production implies a severe radiation environment, up to 1010 n/cm2/s and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$5\times 109\,\,\gamma $ </tex-math></inline-formula> /cm2/s [secondary gammas mainly due to (n, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\gamma $ </tex-math></inline-formula> ) reactions], inside Torus Hall building (THB) and outside the THB nearby the penetrations. DTT will be equipped with several power and signal electronic devices: 1) critical electronics that play a crucial role for the machine functioning in order to avoid failure, loss of plasma stability, and damages on the structural components (e.g., energy supply to the toroidal field, poloidal field, and central solenoid superconductive magnets); and 2) noncritical electronics, used to power diagnostics and systems whose failures do not compromise the machine operations, but are essential for scientific and engineering exploitation programs. Neutronics and shielding studies of electronics are fundamental for protecting the main components of the tokamak and to avoid the failure of the critical systems, as well as for the evaluation of the positioning and functioning of the noncritical devices and the selection of radiation-hard electronics. In this work, 3-D shielding studies for the DTT electronics devices, carried out with MCNP Monte Carlo code, are presented and discussed as well as the effectiveness of the proposed arrangement and mitigation solutions.

Nuclear Analyses for the Assessment of the Loads on the ITER Radial Neutron Camera In-Port System and Evaluation of Its Measurement Performances

The radial neutron camera (RNC) is a key ITER diagnostic system designed to measure the uncollided 14- and 2.5-MeV neutrons from deuterium–tritium (DT) and deuterium–deuterium (DD) fusion reactions, through an array of detectors covering a full poloidal plasma section along collimated lines of sight (LoS). Its main objective is the assessment of the neutron emissivity/ <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\alpha $ </tex-math></inline-formula> source profile and the total neutron source strength, providing spatially resolved measurements of several parameters needed for fusion power estimation, plasma control, and plasma physics studies. The present RNC layout is composed of two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module (DSM) of ITER Equatorial Port 1 (EP01): the ex-port subsystem and the in-port one. The ex-port subsystem, devoted to the plasma core coverage, extends from the Port Interspace to the Bioshield Plug: it consists of a massive shielding unit hosting two sets of collimators lying on different toroidal planes, leading to a total of 16 interleaved LoS. The in-port system consists of a cassette, integrated inside the port plug DSM, containing two detectors per each of the six LoS looking at the plasma edges. The in-port system must guarantee the required measurement performances in critical operating conditions in terms of high radiation levels, given its proximity to the plasma neutron source. This article presents an updated neutronic analysis based on the latest design of the in-port system and port plug. It has been performed by means of the Monte Carlo MCNP code and provides nuclear loads on the in-port RNC during normal operating conditions (NOC) and inputs for the measurement performance analysis.

Feasibility study of using the NUR research reactor for a BNCT installation and Monte Carlo optimization of a BSA

The aim of this work was devoted to study the feasibility of using the Algerian research reactor NUR for Boron Neutron Capture Therapy (BNCT) applications. To achieve this aim, an analysis of parameters and operational safety limits of the current core configuration X-1 with a removal of the thermal column was carried out. Furthermore, this study was followed by optimization calculations of the Beam Shaping Assembly (BSA) in the channel N∘5 using the Monte Carlo code MCNP5 simulations with a highlight of a new moderating material (Modified Fluental). Subsequently, a dosimetric study was performed using the MCNP5 code and the Snyder Head phantom by taking a Boron concentration of 65 ppm and 18 ppm in tumor and healthy tissue respectively. The in-phantom figures-of-merit have been calculated and provided promising results.

Neutronic investigation of a VVER-1200 (Th-233U)O2 fuel assembly with protactinium oxide as a burnable absorber coated on the outer surface of the fuel rods

ABSTRACT Due to the obvious excess reactivity associated with using 233U as a fissile material, fresh fuel must include a burnable absorber (BA) to reduce excess reactivity. As a result, the objective of this study is to look into the use of protactinium oxide (PaO2) as a thin BA layer coating on the outer surface of the fuel rod for the VVER-1200 assembly, which is fueled by (Th-233U)O2. Through neutronic calculation, the thicknesses of PaO2 as a BA were determined using the Monte Carlo code MCNP6. For comparison, UO2 and (Th-233U)O2 assemblies without absorbers were used as references. The study suggests that using PaO2 with a thickness ranging from 0.11 to 0.13 mm can reduce the early excess reactivity of the (Th-233U)O2 assembly while also increasing the discharge burnup by an average of 2.1 GWd/ton when compared to the UO2 assembly. Furthermore, it was found that the selected (Th-233U)O2 + PaO2 assemblies exhibit negative fuel temperature and boron worth coefficients. However, these assemblies exhibit positive moderator temperature coefficients. Neutron spectra, radial power distribution, and fission product poisons were also evaluated to understand the viability of using PaO2 as a BA for reactivity control in the VVER-1200 assembly under consideration.

Conceptual design of a 3 MWth yttrium hydride moderated heat pipe cooled micro reactor

Micro reactors are gaining substantial interest worldwide due to their capabilities to deliver power to remote and decentralized areas and their long fuel cycles, making them more attractive than other energy sources. This paper presents a conceptual design and steady-state neutronics and thermal-hydraulics analyses results for a heat pipe cooled yttrium hydride moderated micro reactor. The reactor was designed for both terrestrial and space application purposes; hence, size and weight were important. The neutronics analysis calculations were performed to evaluate key safety parameters such as control devices reactivity worth, temperature reactivity coefficients, and power distribution of the reactor core. Burnup calculations were performed using Monte Carlo code MCNP6.2 and verified against Serpent code results. The thermal-hydraulics analyses calculations were performed to evaluate the temperature distribution in a representative unit cell in the reactor core using the STAR CCM+ multi-physics computational fluid dynamics (CFD) software by utilizing the power distribution obtained from neutronics calculations. It was demonstrated that this reactor could be operated safely for more than 11 years, generating 3 MW of thermal power.

Assessment of nuclear data libraries for fast reactor systems with MCNP and NDaST

The paper presents results for multiplication factors of reactor physics experiments with fast spectrum systems, representative of liquid metal fast reactors, documented in the IRPhEP Handbook. The calculations were performed with the Monte Carlo code MCNP 6.1, using continuous-energy data based on ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.1.2, JEFF-3.2, and JEFF-3.3. Considerable discrepancies were observed between the results obtained with the different data libraries. In comparison with the benchmark values of the experiments, the ENDF/B-VII.1 results show the best agreement, while for the JEFF-3.3 results, the deviations are largest. In addition, the sensitivity tool NDaST, using linear perturbation theory, was applied to identify the most important microscopic cross sections responsible for the differences of the latest library revisions ENDF/B-VIII.0 and JEFF-3.3 as compared to ENDF/B-VII.1. For JEFF-3.3 vs ENDF/B-VII.1, where the deviations are overall largest, the main contributors are U-238n, γ and, for systems containing plutonium, Pu-239 fission and total neutron multiplicity.

Analytical Discrete-Ordinates Solutions for Improved 1D Models of Particle Transport in Ducts with Wall Migration

Analytical discrete-ordinates (ADO) solutions are developed for two improved one-dimensional (1D) models of particle transport in ducts that include wall migration. One of the studied models is based on an approximation of the transverse and azimuthal dependencies of the angular flux in terms of two basis functions, while the other uses three. Particle migration in the wall is modeled by an exponential displacement kernel. Numerical results for duct reflection and transmission probabilities are reported for various test cases, including some cases of thermal-neutron transport through iron, concrete, and graphite ducts that have been defined and studied by other authors. A comparison is performed with a set of published results comprising realistic results obtained with the Monte Carlo code MCNP and results from a numerical implementation of the discrete-ordinates method for the model based on three basis functions. To resolve some issues raised during the comparison process, a numerical discrete-ordinates solution of the problem has also been implemented in the course of this work.

Radiation shielding analysis of the HI-STORM FW storage cask

This paper presents radiation shielding analysis of the HI-STORM FW storage cask designed by the Holtec International, Inc. for the Nuclear Power Plant Krsko Spent Fuel Dry Storage (SFDS) project. The first part of the analysis is addressing the problem of a particle source generation. For each spent fuel assembly (SFA) neutron and gamma source intensities and spectra were calculated using ORIGEN-S module of the SCALE 6.2.3 code package. The second part of the analysis is considering neutron and gamma dose rates calculation, for which hybrid shielding methodology based on Monte Carlo (MC) code MCNP6.2 and ADVANTG3.0.3 code was used. The dose rates for the reference case included the contributions of fuel neutron, fuel gamma, secondary gammas from (n, γ) reactions, as well as gammas from activated SFA hardware components due to Co-60 presence. Expectedly, the results have shown that the dose rates around the cask have nonuniform behavior due to a specific design of the cask. The heavy concrete body of the cask presents a good attenuator for all types of radiation, however, a considerable streaming through the top and bottom ventilation openings was noticed. It was also observed that an important role was played by the isotope Co-60 which emits two lines of high energy deep penetrating gammas. The convergence of dose rates and associated relative uncertainties during MC calculations have shown that significantly more histories were required for gamma calculations compared to neutron calculations. To investigate the impact of different SFA content in the casks on the dose rates, shielding analysis was made for 16 casks that will be loaded in the first campaign. For comparison, the calculations of a single HI-STORM FW cask were additionally performed using MAVRIC module of the SCALE 6.2.3 and the obtained differences in external dose rates of 20–30% are expected. The calculation cases showed the differences in casks doses due to different casks loading within loading plan criteria, reduction provided by heavy concrete in cask body (more important for gammas), importance of careful design of lower and upper casks openings to keep increase of dose rates, compared to closed configuration, within one order of magnitude, and compromise needed when describing details of intensity and spectrum of SFA sources. The increased calculation mesh density was used to show that selected discretization is appropriate, and the limited influence of storage pad thickness was demonstrated by performing calculation with thick concrete slab corresponding to Dry Storage Building (DSB) floor foundation thickness.

Criticality Safety Analysis of Spent Fuel Storage Pool for NPP Mochovce using MCNP5 Code

The paper presents results of nuclear criticality safety analysis of spent fuel storage and handling for the 1st and 2nd unit of NPP Mochovce. Spent fuel storage pool (compact and reserve grid) and T-12 transport cask were modeled using the Monte Carlo code MCNP5. Conservative approach was applied and calculation of max eff k values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal max eff k values. The requirement of current safety regulations to ensure 5% subcriticality was met except some especially conservative cases.

Effective dose rate conversion coefficients for photons from radionuclides calculated using the Monte Carlo method and adult voxel phantoms

In radiation exposure situations, knowledge of the equivalent or effective dose is required to determine whether the individuals are working within safe limits. However, in most situations, doses cannot be measured directly, and tissue or organ absorbed doses and effective doses are determined computationally to help assess stochastic risks. This study incorporates voxel-based whole body models of ICRP (adult reference male and female phantoms) into the Monte Carlo code MCNP 6.1 to simulate radiation interactions inside the human body due to the photons emitted from 1179 different radioisotopes. Absorbed doses in tissues or organs were computed from simulations where a radioactive point source was placed 1 m away at 1 m above the ground for eight different source orientations with respect to each phantom. The results were subsequently multiplied by the radiation weighting factors (WR = 1 for photons) and the tissue weighting factors in accordance with the methodology of ICRP 103 to obtain organ equivalent doses and effective doses. The calculated effective dose for each radioisotope and source orientation was later divided by unit source activity and unit exposure time to convert to effective dose rate conversion coefficient (EDCC). This novel quantity is proposed in this study for more realistic dose assessments. The results show that the anterior-posterior sources led to the highest EDCC values for all the radioisotopes. In addition, the right posterior orientations yield the lowest EDCCs for most of the radioisotopes. Under all source orientations, EDCC values were highest for Cf-254 isotope and lowest for Sm-151. EDCC to traditional air kerma rate constant (AKRC) ratios were also calculated to compare these two quantities. For all source orientations and most radioisotopes, this ratio was less than one and its largest value was observed for the anterior-posterior source. The results obtained in this study are based on effective doses from photon emitting radioisotopes developed for whole body models under various source orientations, thus should more accurately represent the stochastic risk to individuals in an external exposure situations.

Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods

This paper presents comparison of the Krško Power Plant simplified Spent Fuel Pool (SFP) dose rates using different computational shielding methodologies. The analysis was performed to estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as a bulk region materials. The pool is divided on the old and new section where the old one has three additional subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year and 5 years). The new section represents the FAs with the cooling time of 10 years. The time dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield calculation is based on imported isotopic activities. The time dependent photon spectrum with total source intensity from Microshield multigroup point kernel calculations was then prepared for two hybrid deterministic-stochastic sequences. One is based on SCALE6.2b3/MAVRIC (Monaco and Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1. code. Even though this model is a fairly simple one, the layers of shielding materials are thick enough to pose a significant shielding problem for MC method without the use of effective variance reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR parameters for the MCNP fixed-source calculation using continuous energy transport. ADVATNG employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FWCADIS methodology. Denovo uses a structured, Cartesian-grid SN solver based on the Koch- Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was our first application of ANDVANTG/MCNP hybrid sequence for this type of calculation and the results where compared to SCALE/MAVRIC sequence which we regularly use for similar calculations. The comparison of gamma dose rates on different point detector locations (central above pool and at the top of pool periphery) showed a good agreement between Microshield (point-kernel) and deterministic-stochastic shielding methodologies for the cylindrical approximation of the pool geometry. More complicated cases for model with multi-source option and for cuboids showed very good agreement between SCALE/MAVRIC and ANDVANTG/MCNP calculations.

Preparation and evaluation of gamma shielding properties of silicon-based composites doped with WO3 micro- and nanoparticles

This study aims to investigate the shielding properties of silicon-based WO3 composites against gamma radiation. The filler particles in micro and nano-size were prepared with a wet chemical synthesis method. Consequently, different weight percentages of micro and nano WO3 composites (5, 10, 15, 20, 30, 50, and 70 wt%) were produced. Structural and morphological characterizations of the composites were investigated using SEM and EDS analyses. Moreover, for the thermal characterization of the nanocomposites and to prove the thermal stability, TGA analysis was carried out. The results indicated that adding WO3 nanoparticles to the silicon matrix decreases the glass transition temperature of the composites. Additionally, shielding properties of the nano and micro composites were investigated using a NaI(Tl) scintillator detector at different photon energies and experimental results verified by MCNP Monte Carlo code and XCOM program. It was illustrated that, especially at lower energies, in both micro-and nanocomposites samples, by increasing the weight percentages of the filler particles, shielding abilities of the composites enhance appreciably.