Monte Carlo Code Research Articles

Design and implementation of HDF5-format neutron nuclear database in RMC code

This study developed a new structured nuclear database to improve the readability and extensibility of the ACE (A Compact ENDF) nuclear database, save disk space, reduce memory usage, and enhance the computational efficiency of the Reactor Monte Carlo (RMC) code. A Python package was developed to store nuclear data in HDF5 format. Compared to the ACE database, the HDF5-format database shows significant improvements: an 80% reduction in disk space for continuous energy neutron data and a 60% reduction for neutron thermal scattering data. The HDF5-format database was implemented in RMC’s criticality calculation mode and validated through VERA benchmark problem 2B, demonstrating perfect agreement with the ACE results in keff and neutron flux counts. The Computational results indicate a 7.7% reduction in memory usage and a 20.1% improvement in computational efficiency with the HDF5-format database. Additional tests show that using the database at a single temperature point reduces memory usage by 5.4% and running time by 13.2%. At two temperature points, memory usage decreases by 35.2% and running time by 18.0%. The new data structure reduces temperature-independent redundant data and improves indexing efficiency, leading to greater savings with more temperature points. This development enhances performance of criticality calculation of RMC and addresses ACE database limitations.

Preliminary neutronic analyses for annular VHTR with different coated fuel particle designs

Abstract The purpose of the present work is to determine the neutronic characteristics of annular cylindrical prismatic Very High Temperature Gas-cooled Reactor (VHTR) with different advanced fuel designs. The study includes detailed neutronic investigations of the eight types of TRISO (containing SiC layer) and TRIZO (containing ZrC layer) fuel particles for annular VHTR core. All calculations were performed using continuous energy Monte Carlo code MVP2.0 and MVPBURN with JENDL4.0 nuclear data library. The obtained results indicated that SiC + PyC and PyC layers in TRISO fuel appeared to be not excellent choice for the annular VHTR core due to its shorter operation year and, lower fuel burnup. The annular cylindrical prismatic core fueled with TRIZO fuel particle shows good neutronic characteristics compared to those designed with other TRISO fuel particles.

Experimental and theoretical study on the physico-mechanical characteristics and radiation shielding capability of hardened alumina sludge waste-cement pastes containing MnFe2O4-nanoparticles

This research investigates the impact of incorporating low-cost MnFe2O4 spinel nanoparticles (MF-NPs) at varying concentrations (0.5, 1, and 2 mass%) on the mechanical and physical properties of blended cement pastes. These pastes were produced by replacing different proportions (5, 10, and 15 mass%) of ordinary Portland cement (OPC) with activated alumina sludge waste (AAS), to promote sustainability. Also, the research examined the gamma radiation shielding effectiveness of certain hardened composites against a 137Cs gamma radiation source with an energy of 661.64 keV using a NaI (Tl) detector (Oxford Model) with 3″ × 3″, amplifier and 16 k multi-channel analyzer. The linear attenuation coefficient (LAC) of all the studded samples were calculated theoretically using a Monte Carlo code MCNP-5 code. The gamma radiation shielding properties were analyzed in depth using a Monte Carlo code MCNP-5 simulation model. The theoretical and experimental results for LAC were found to be in complete agreement. Phy-X/PSD software was applied to estimate the mass attenuation coefficient (MAC) for gamma radiation at various energies, as well as the effective atomic number (Zeff), mean free path (MFP), half-value layer (HVL), and tenth-value layer (TVL). The findings demonstrated that the addition of 0.5% MnFe2O4 nanoparticles (MF-NPs) to blended cement pastes exhibited the best physical and mechanical characteristics, as well as the most effective gamma radiation shielding.

In silicodosimetry for a prostate cancer treatment using198Au nanoparticles.


To estimate dose rates delivered by using radioactive198Au nanoparticles for prostate cancer nanobrachytherapy, identifying contribution by photons and electrons emmited from the source.
Approach:
Utilizingin silicomodels, two different anatomical representations were compared: a mathematical model and a unstructured mesh model based on the International Commission on Radiological Protection (ICRP) Publication 145 phantom. Dose rates by activity were calculated to the tumor, and nearby healthy tissues, including healthy prostate tissue, urinary bladder wall and rectum, using Monte Carlo code MCNP6.2.
Main results:
Results indicate that both models provide dose rate estimates within the same order of magnitude, with the mathematical model overestimating doses to the prostate and bladder by approximately 20% compared to the unstructured mesh model. The discrepancies for the tumor and rectum were below 4%. Photons emmited from the source were defined as the primary contributors to dose to other organs, while 97.9% of the dose to the tumor was due to electrons emmited from the source.
Significance:
Our findings emphasize the importance of model selection in dosimetry, particularly the advantages of using realistic anatomical phantoms for accurate dose calculations. The study demonstrates the feasibility and effectiveness of198Au nanoparticles in achieving high dose concentrations in tumor regions while minimizing exposure to surrounding healthy tissues. Beta emissions were found to be predominantly responsible for tumor dose delivery, reinforcing the potential of198Au nanoparticles in localized radiation therapy. We advocate for using realistic body phantoms in further research to enhance reliability in dosimetry for nanobrachytherapy, as the field still lacks dedicated protocols.&#xD.

The dosimetric accuracy of a commercial model-based dose calculation algorithm in modeling a six-groove direction modulated brachytherapy tandem applicator

Objective.With advancements in high-dose rate brachytherapy, the clinical translation of intensity modulated brachytherapy (IMBT) innovations necessitates utilization of model-based dose calculation algorithms (MBDCA) for accurate and rapid dose calculations. This study uniquely benchmarks a commercial MBDCA, BrachyVision ACUROSTM(BVA), against Monte Carlo (MC) simulations, evaluating dose distributions for a novel IMBT applicator, termed as thesix-grooveDirection Modulated Brachytherapy (DMBT) tandem, expanding beyond previous focus on partially shielded vaginal cylinder applicators, through a novel methodology.Approach.The DMBT tandem applicator, made of a tungsten alloy with six evenly spaced grooves, was simulated using the GEANT4 MC code. Subsequently, two main scenarios were created using the BVA and reproduced by the MC simulations: 'Source at the Center of the Water Phantom (SACWP)' and 'Source at the Middle of the Applicator (SAMA)' for three cubical virtual water phantoms (20 cm)3, (30 cm)3, and (40 cm)3. A track length estimator was utilized for dose calculation and 2D/3D scoring were performed. The difference in isodose surfaces/lines (i.e. coverage) at each voxel,ΔDIsodose Levels/Lines, was thus calculated for relevant normalization points (rref).Results.The coverage was comparable, based on 2D scoring, between the BVA and MC isodose surfaces/lines for the region of clinical relevance, (i.e. within 5 cm radius from the source) withΔDIsodose Lines(rref: 1 cm from the source) falling within 2% for the two scenarios for all phantom sizes. For the phantom (20 cm)3,ΔDIsodose Levels(3D scoring) recorded the range [-3.0% +6.5%] ([-7.4% +7.3%]) for 95% of the voxels of the same scoring volume for the SACWP (SAMA) scenario.Significance.The results indicated that the BVA could render comparable coverage as compared to the MC simulations in the region of clinical relevance for different phantom sizes.ΔDIsodose Linesmay offer an advantageous metric for evaluation of MBDCAs in clinical setting.

Insights into calculating Reference Discontinuity Factors with Serpent Monte Carlo code

This study explores the calculation of Reference Discontinuity Factors (RDFs) using the Serpent Monte Carlo code, focusing on the methodology and potential pitfalls. In two-step reactor analyses, consistently generated RDFs are crucial for aligning homogeneous nodal diffusion results with the reference heterogeneous transport solution. However, the Serpent internal diffusion solver, based on the Analytic Function Expansion Nodal (AFEN) method, may not be compatible with other nodal methods such as the Nodal Expansion Method (NEM). Additionally, the solver can suffer from instabilities, particularly in multi-group calculations, leading to erroneous RDFs. Despite these challenges, Serpent can generate the necessary raw data for RDF calculation, which can be accurately processed using external diffusion solvers. Two numerical examples − a 1D fuel-reflector model and a 2D SMR core model − illustrate the effects of consistent and inconsistent RDFs on simulation accuracy. The study emphasizes the importance of using compatible diffusion solvers and thoroughly assessing RDFs to avoid errors in reactor simulations.

Advancements to generalized equivalence theory for preserving cross-sections of whole core simulations

The conventional two-step approach, without the use of an equivalence method, can introduce significant error for the simulation of non-LWRs (Light Water Reactors), especially fast reactors, due to the large leakage and high anisotropic neutron distribution. To improve upon the accuracy of the two-step approach, a 3D whole core model is simulated with a transport method to generate region-wise homogenized cross-sections (XS). These XS can then be used in a diffusion whole core solver during the second step to extend the application to cycle and transient analysis. Discontinuity factors (DFs) are then introduced to improve the accuracy during the simulation of the second step. With properly generated DFs from Generalized Equivalence Theory (GET), the region-averaged solutions from the first transport step can then be reproduced by the second step diffusion solver. The Monte Carlo method was selected to perform the whole core transport simulation to generate region-wise XS. However, simulating whole core problems with Monte Carlo may result in poor statistics near the peripheral region especially for partial current tallies. This paper introduces an advancement to GET to reproduce region-wise solutions for select regions when the reaction rates and surface currents have good statistics in these regions but poor statistics in other regions. A reference high-fidelity model was constructed using the Serpent 2 Monte Carlo code based on the EBR-II benchmark evaluation and verification was carried out using the TriPEN-4 method in PARCS. The results show that it is possible to reproduce the exact eigenvalue and power distributions of whole core problems in a feasible manner.

The FLUKA Monte Carlo simulation of the magnetic spectrometer of the FOOT experiment

The FOOT experiment of INFN is devoted to the measurement of the nuclear fragmentation double differential cross sections useful for the improvement of calculation models adopted in hadrontherapy and radioprotection. A detailed Monte Carlo simulation of the FOOT magnetic spectrometer has been implemented in order to optimize the design and to guide data analysis. This task has been accomplished by means of the FLUKA Monte Carlo code. The input files of the FLUKA simulations are created from the software framework of the experiment, in order to have a consistent generation and description of geometry and materials in both simulation and data analysis. In addition, this ensures the possibility of processing both simulated and real data with the same data analysis procedures. Databases containing specific parameters describing the setup employed in each different data taking campaign are used. A customized event-by-event output of the Monte Carlo code has been developed. It can be read out by the general software framework of FOOT, enabling access to the generation history of all particles in the same event. This output structure therefore gives the possibility to perform a detailed analysis and study of all relevant processes, allowing the detailed tracking reconstruction of all individual particles. Examples of results are presented.

Development of an EGSnrc multi-leaf collimator component module and treatment head model for a low-field MRI linear accelerator.

Monte Carlo (MC) modeling of MR-guided radiotherapy (MRgRT) treatment machines enables the characterization of photon/electron interactions in the presence of a magnetic field. The EGSnrc MC code system is a well-established system for radiation dose calculations. The multi-leaf collimator (MLC) component modules presently available within the EGSnrc MC code system do not include a model of the double-focused MLC available on a low-field (0.35T) MRI linear accelerator (MR linac). Here we developed and validated a new component module (CM) for the low-field MRgRT MLC using the EGSnrc/BEAMnrc/DOSXYZnrc code system. We performed detailed modeling of the treatment head and validated the model using measurements and calculations from the vendor-specific treatment planning system (TPS). The detailed geometry of the low-field MR linac MLC and other treatment head structures were modeled using BEAMnrc. Comparisons of DOSXYZnrc simulated dose against measurements and the low-field MR linac TPS for a variety of AAPM TG-53 task group report suggested square and shaped fields, as well as a step-and-shoot intensity-modulated radiotherapy (IMRT) plan, are presented. Our model agrees with both measured and TPS calculated data on average within 2%/2mm (dose/DTA) criterion for square field profiles. Output factors agreed within 1% for field sizes down to 2.49×2.49cm2 and within 2% of TPS data for the smallest field size of 0.83×0.83cm2. Shaped field and IMRT MC calculations agreed with measured and TPS data such that the gamma pass rates (3%/2mm) were 99.5% and (3%/3mm) 96.2%, respectively. We developed and validated an MLC CM (SYNCVRMLC) for the low-field MR linac using the EGSnrc MC code systems. This new CM will facilitate MC computation of fluence and dose distributions using BEAMnrc/DOSXYZnrc for patients treated on the low-field MR linac.

Dynamics of particle-particle correlations and the ridge effect in proton-proton collisions

In high-energy particle physics, the study of particle-particle correlations in proton-proton and heavy-ion collisions constitutes a pivotal frontier in the effort to understand the fundamental dynamics of the strong force. To the best of our knowledge, we employ for the first time the Balitsky-Fadin-Kuraev-Lipatov dynamics implemented in a Monte Carlo code in momentum space to compute final-state correlations in proton-proton collisions. Our present work aims to investigate whether the particular dynamics of the high-energy limit of QCD can contribute to the long-range rapidity correlations and the enigmatic ridge effect in proton-proton collisions. Published by the American Physical Society 2024

ITER full model in MCNP for radiation safety demonstration

The development of nuclear fusion as a safe and virtually limitless power source is receiving growing attention in the context of looming energy crisis and climate change. ITER project stands as the flagship international initiative and is advancing steadily. The construction of the Tokamak Complex is nearly finished, and the assembly of core components has begun on site. Simultaneously, the design is being finalized, and the safety case is becoming more concrete. Current approaches to radiation safety demonstration using 3D nuclear analysis with the Monte Carlo code MCNP require sophisticated artifacts to sew together simulations in separate models for the Tokamak and the rest of the facility. This results in cumbersome studies and, consequently, challengeable conclusions. To address this issue, we have built the an integral MCNP model of the ITER facility: the ITER full model. Along with improvements to the D1SUNED code, we illustrate its computational practicality and pertinence in two meaningful simulations for ITER safety case. This work represents the culmination of a two-decade-long effort of ITER modelling aiming to demonstrate adequate radiation safety. Beyond supporting the remaining design tasks, this model simplifies the corresponding 3D nuclear analysis and improves the robustness of the ITER safety case.

Bulk, overlap and surface effects of swift heavy ions in CeO2

Formation of tracks of swift heavy ions decelerating in the electronic stopping regime in CeO2 was studied, combining the Monte Carlo code TREKIS with molecular dynamics. We show that strong lattice disordering (melting) followed by structure recovery form finally a damaged ion track consisting of a discontinuous crystalline region in CeO2. Normal ion impacts result in appearance of spherical crystalline hillocks on CeO2 surface. The solid-vacuum interface strongly suppresses the recrystallization of the near-surface layers, forming conically shaped tracks with several tens of nanometers lengths. Grazing ion irradiation induces intensive material expulsion from the surface forming finally grooves surrounded by nanohillocks. The processes of surface nanostructures formation is similar to those observed previously in CaF2 which has the similar crystalline structure, however requires much longer recrystallization time. Recent experimental data confirm the simulation results.

Investigation of photon interaction parameters in Human Body Tissues using GAMOS, FLUKA, and XCOM Studies

This study aims to evaluate different photon interaction parameters of human body organs at nine photon energies (0.021 keV to 1.25 MeV) using GAMOS and FLUKA Monte Carlo codes. The Mass attenuation coefficients, Mean Free Path, Half Value Layer, Tenth Value Layer, and Mass absorption coefficients of the breast, lung, kidney, pancreas, liver, eye lens, thyroid, brain, ovary, heart, large intestine blood, skin, spleen, muscle, and cortical bone were investigated and compared against the standard values available from the tables of XCOM database. This comparison demonstrated a good agreement, with all calculated values deviating less than 6%. In addition, the values obtained with the GAMOS code were similar to those obtained using the FLUKA code. These comparisons indicate very good agreement and revealed that both FLUKA and GAMOS codes can be used as a computational tool for determining the photon interaction parameters of any material and also different energies, whose data are not available in the literature.

Influence of TiO2 content on the radiation shielding properties of the La2O3-B2O3-Gd2O3-Nb2O5-ZrO2-SiO2 glasses

This paper aims to study the radiation shielding properties of lanthanide glasses, according to the formula xTiO2-51La2O3-(24-x) B2O3–8Gd2O3-8Nb2O5-6ZrO2-3SiO2 (x= 0, 4, 8, 12, 16, wt.%). Using FLUKA Monte Carlo code, the mass attenuation coefficients (MAC), half-value layers (HVL), and effective atomic numbers (Zeff) of the lanthanide glasses were estimated at medical diagnostic energies (between 20 and 150 keV). The MACs of the glasses are between 0.5183 and 24.407 cm2/g for 0Ti, 0.5215-24.788 cm2/g for 4Ti, 0.5193-25.161 cm2/g for 8Ti, 0.5163-25.529 cm2/g for 12Ti, and 0.5183-25.916 cm2/g for 16Ti. These results are consistent with the Phy-X theoretical database (with a percentage difference below 3 %). The lanthanide glasses showed good photon shielding ability compared to lead concrete, and RS-360 & RS-253-G18 commercial glasses, commonly used shielding materials. In this work, 16Ti possesses the highest, lowest, and highest values of MAC, HVL, and Zeff, respectively, at the various energies investigated, which implies that the 16Ti sample has better shielding performance. All in all, this work demonstrated that adding TiO2 to the glass samples could provide preferable shielding features.

Optimization of Fuel Enrichment and Reactor Core Pitch Dimension in Hydride Microreactor using NSGA-II Method and OpenMC Calculation

Abstract This study focuses on the design optimization of a hydride microreactor, aimed at supporting the equitable development of electricity in remote regions of Indonesia. Its compact design, inherent safety properties, and independence from the conventional electric grid make this microreactor concept suitable for electrifying even the most inaccessible areas. Current research investigates various aspects of the hydride microreactor design, employing a standard reactor core pitch dimension of 2.4 cm and fuel enrichment of 12%. It is crucial to operate the reactor using specific concentrations of Gadolinia burnable poison to enable a shutdown in the event of a stuck rod. By optimizing only the pitch dimension and fuel enrichment, the reactor can be simulated relatively quickly without the need for burnup calculations. Ideal configuration aims for lower enrichment, an extended fuel cycle, and a relatively low power peaking factor (PPF). The optimization process utilizes the Non-dominated Sorting Genetic Algorithm – II, with the neutron multiplication factor (keff) and PPF as primary objectives. Fuel cycle and feedback reactivity are the sub-objectives that are used to identify the optimal solution from the Pareto front. The calculation process of keff, PPF, fuel cycle, and feedback reactivity coefficients for specific pitch dimensions and fuel enrichment are done by utilizing the OpenMC Monte Carlo code. The optimal configuration is obtained through four generations of NSGA-II, featuring a reactor core pitch of 2.1 cm and fuel enrichment of 10.4%. This configuration yields a keff value of 1.0712049 and a PPF value of 2,0728 at the Beginning of Life (BOL). Remarkably, this configuration enables safe operation without using burnable poison for a continuous period of 12 years, generating 1 MW of thermal power. Emphasizing the reduction of fissile material is also essential to enhance the inherent safety of the reactor.

Benchmark Analysis of a Prismatic Type-high Temperature Gas-cooled Reactor with the ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 Nuclear Data Libraries

In this work we performed a benchmark analysis of the High Temperature Engineering Test Reactor (HTTR) fully-loaded start-up critical core with a 30-bundle loading configuration using the Monte Carlo code Serpent 2 with the recently released nuclear data libraries ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0. The purpose of the work is to reveal the impacts of using different ENDF nuclear data libraries on neutronics calculations of a prismatic high temperature gas-cooled reactor (HTGR). The benchmark results obtained with Serpent 2 were compared against the available experimental values and those attained by different computer codes (MCNP5 and SCALE) using the nuclear data library ENDF/B-VII.0. The comparative results showed good agreement between Serpent 2, the experimental values, MCNP5 and SCALE. The results also exhibited a notable difference between using ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 in predicting the neutronics parameters of the HTTR that suggests further investigation in future work.

ESTIMACIÓN TEÓRICA Y COMPUTACIONAL DEL STOPPING POWER PARAELECTRONES Y PROTONES DE UN DOSÍMETRO PAGAT INFUNDIDO CONNANOPARTÍCULAS DE ORO EN RANGO ENERGÉTICO DE APLICACIONES TERAPÉUTICAS

The present work focuses on the behavior of the stopping power of a polymer gel dosimetry system, known as PAGAT,infused with high atomic number nanoparticles formed by gold, with a 1 % mass concentration. The study is carried out for different energies of electron and proton beams within ranges of therapeutic applications, first by theoretical assessment, and then by simulation with Monte Carlo codes: PENELOPE in the case of electrons, and FLUKA in the case of protons. The results obtained analytically show promising behavior for gold nanoparticle infused tissue-equivalent systems, while the simulation results show better correspondence with models that consider the net path traveled, instead of models based on the thickness of the samples as typically done in experimental determinations.

Optimized Dropping and Rolling and Virtual Surface Method for High Packing Fraction of Dispersed Fuel Particles in Annular Container

Because of their favorable thermal conductivity and commendable fuel performance, dispersed fuel elements have been widely adopted in research reactors. However, the geometric complexity resulting from the high packing density of dispersed fuel poses challenges to current geometric modeling methods. Various implicit and explicit modeling techniques have been implemented in Monte Carlo codes. Implicit methods struggle to achieve the correct packing fraction and, therefore, lack accuracy. In this paper, we introduce the Optimized Dropping and Rolling and Virtual Surface method for precise and efficient geometric modeling of such structures. Simultaneously, we apply this novel method to the packing process of an annular container. Test results across various packing fraction cases demonstrate that our modified Optimized Dropping and Rolling method significantly enhances packing efficiency, surpassing the Distinct Element Method hundreds of times.

Application of the PSI cycle check-up methodology for full-core Monte Carlo simulations verified using pin power estimates and validated for hot zero power conditions

Monte Carlo (MC) based neutron transport codes provide high-fidelity numerical solutions to problems beyond the modelling capabilities of conventional core simulation methods. However, performing core-follow calculations with MC codes remains challenging due to the necessity of incorporating thermal-hydraulic feedback and generating evolved fuel composition for the cycle of interest by taking into account the entire irradiation history of loaded fuel assemblies over previous cycles. Nevertheless, at PSI, the neutronic version of a cycle check-up methodology (CHUP) is under development to address this challenge. This methodology involves extracting operating conditions and nuclide compositions from validated reference deterministic core-follow models (CASMO5/SIMULATE5/SNF), subsequently generating MC neutron transport models for selected operating points. This article presents the verification and validation performed for a hot zero power operating condition of a Swiss pressurised water reactor using the updated CHUP methodology, which automates information extraction from reference core models to minimise human intervention. The MC models were verified by assessing their deviation from criticality, consistently found to be supercritical within the [10,60] pcm range. Additionally, radial relative power distribution verification yield deviations in the [−5,4] % range compared to validated nodal results. Finally, additional validation was performed using start-up measurements from two successive cycles. Eight MC models were found to be subcritical, on average by 51 pcm, with a standard deviation of 25 pcm.

Estimation of Dose Perturbation Due to the Presence of Metal Hip Prostheses in Radiotherapy with Electron and Photon Beams: A Monte Carlo Study

Purpose: The impact of various hip prosthesis materials on the amount of dose perturbation generated by 9 MeV electron and 18 MV photon beams during pelvic radiation therapy is analyzed in this research. Materials and Methods: The Varian 2100 C/D LINAC head for the electron (9 MeV) and photon (18 MV) modes and a water phantom with a realistic hip prosthesis were modeled using the Monte Carlo (MC) code; MCNPX (Ver. 2.6.0) and were benchmarked for measurement. Four different materials, including Cobalt–Chromium–Molybdenum-alloy (CCM), Stainless Steel (SS), Titanium, and Titanium-alloy (Ti-alloy) were evaluated. The changes in electron and photon fluences and dose perturbations due to the presence of the hip prostheses were investigated. Results: An increased dose of 13.29%, 13.77%, 6.16%, and 5.93% for 9 MeV and 30.43%, 33.05%, 10.89%, and 11.27% for 18 MV was calculated for Ti-alloy, Ti, CCM, and SS prosthesis, respectively. At 0.5 mm distance from the prosthesis, the Electron Backscatter Factor (EBF) of 1.31, 1.33, 1.15, and 1.14 for 9 MeV and 1.32, 1.34, 1.11, and 1.12 for 18 MV was calculated for Ti-alloy, Ti, CCM, and SS prosthesis, respectively. Dose perturbation is higher at a near distance from the prosthesis; by reducing the distance from the Ti-alloy prosthesis (1.5 to 0.5 mm), an increased dose of 7.70% and 5.54% resulted in 18 MV and 9 MeV, respectively. The dose decreases of up to 21% behind the hip prosthesis were calculated for 18 MV. Conclusion: It is essential to consider the dose perturbation due to the presence of a hip prosthesis to achieve the optimal treatment planning in radiotherapy.