Monte Carlo Photon Transport Code Research Articles

A new Excel-based Monte Carlo transport code for simulating energy response spectra and efficiencies in gamma-ray detection materials

This study presents and assesses a new, user-friendly, and Excel-based Monte Carlo photon transport code, specifically designed to simulate Gaussian energy broadened response spectra of gamma-ray detector materials. This vectorized code utilizes the EPICS2017 photoatomic data library, with unionized energy grid construction, to facilitate precise and efficient photon transport simulations in any user-defined materials composition. Performance benchmarks were conducted using PHITS 3.31, with both basic and detailed models of a gamma-ray scintillator system, and detection crystals of NaI(Tl) and LaBr3(Ce). These benchmarks spanned a wide array of photon source monoenergetic emissions. Detector model efficiency was determined through automated peak area analysis, employing algorithms from the Ortec GammaVision software. The results indicated only minor spectral differences between the Excel-based code and the PHITS models, particularly within the range of practical gamma-ray energies and near-zero source-detector distances. Efficiency curves demonstrated good agreement between the Excel-based code and PHITS models, especially within the energy range of approximately 200 keV to a few MeV, with increasing agreement as detector crystal size increased. Comparisons were conducted with the existing data in literature at source-detector distances of 2 cm and 5 cm, using a detector crystal size of 2”. These comparisons support the code’s validity and its implementation of a basic geometry. This new Excel-based code appears as a promising tool with potential applications in radiation detector characterization and materials research.

Effect of the various tritium breeding materials on the tritium breeding ratio in ARC fusion reactor

This study deals with TBR (Tritium Breeding Ratio) performance of the four different types of molten salt (FLiBe, FLiNaK, FLiNaBe, FLiNaRb) which has four different Li-6 enrichment ratios (0%, 30%, 60%, 90%) Moreover, two different vacuum vessel materials (SiC composite and V4Cr4Ti alloy) have been used in ARC reactor geometry. In models with TBR ≥1.05, it has been desired to examine the sustainability of the tritium production of the reactor by removing the neutron multiplication layer from the design without using the toxic Beryllium element. The neutronic performance studies have been examined in OpenMC, which is an open-source Monte Carlo neutron and photon transport code. The TBR values of the different models obtained and the change of these values according to Li-6 enrichment were examined. Obtaining result showed that the FLiBe molten salt and V4Cr4Ti alloy vacuum vessel wall material could be the best choice for the ARC reactor.

Current Status and On-Going Development of VTT’s Kraken Core Physics Computational Framework

The Kraken computational framework is a new modular calculation system designed for coupled core physics calculations. The development started at VTT Technical Research Centre of Finland in 2017, with the aim to replace VTT’s outdated legacy codes used for the deterministic safety analyses of Finnish power reactors. In addition to conventional large PWRs and BWRs, Kraken is intended to be used for the modeling of SMRs and emerging non-LWR technologies. The main computational modules include the Serpent Monte Carlo neutron and photon transport code, the Ants nodal neutronics solver, the FINIX fuel behavior module and the Kharon thermal hydraulics code, all developed at VTT. The core physics solution can be further coupled to system-scale simulations. In addition to development, significant effort has been devoted to verification and validation of the implemented methodologies. The reduced-order Ants code has been successfully used for steady-state, transient and burnup simulations of PWRs with rectangular and hexagonal core geometry. The Ants–Kharon–FINIX code sequence is actively used for the core design tasks in VTT’s district heating reactor project. This paper is a general overview on the background, functional description, current status and future plans for the Kraken framework. Due to the short history of development, Kraken has not yet been comprehensively validated or applied to full-scale core physics calculations. A review of previous studies is instead provided to exemplify the practical use.

Development and preliminary verification of a Monte Carlo photon transport code IMPC-Photon

The Monte Carlo photon transport code IMPC-Photon is developed for serving the software development of designing advanced nuclear energy system such as Accelerator Driven Advanced Nuclear Energy System (ADANES) in Chinese Academy of Sciences. For photon physics, four main photon-atomic interactions and three secondary processes including atomic relaxation, electron-positron annihilation, and thick-target bremsstrahlung are implemented. For code calculations, the unified energy grid method is applied to improve the speed of searching for the cross section. Under the same computational conditions, the unified energy grid method is more time-efficient and the search efficiency is significantly improved. In addition, the “point gamma-ray source in infinite medium” and “gamma-ray skyshine experiment” benchmarks are employed to verify the reliability of the code. The results show that IMPC-Photon has a good agreement with reference codes and experimental measurement, which indicate that IMPC-Photon is reliable and credible for photon transport problem.

Analysis of BEAVRS two-cycle benchmark using RMC based on full core detailed model

With the increasing demands of high fidelity neutronics analysis and the development of computer technology, Monte Carlo method is becoming more and more attractive especially in criticality analysis of initial core and shielding calculations, due to its advantages of flexible geometry modeling and use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with different physics and feedback interactions and coupling. To perform the high fidelity multi-physics simulations of real reactors or benchmark calculations such as two-cycle BEAVRS benchmark based on measurement data of a practical nuclear power plant, several factors must be considered such as large scale detailed depletion, thermal-hydraulics feedback, on-the-fly nuclear cross section processing, criticality search and inter-cycle refueling. In this paper, the abilities mentioned above for multiple burnup cycles simulations in Hot Full Power condition of PWR full core have been developed in continuous-energy Reactor Monte Carlo neutron and photon transport code RMC, which is developed by Department of Engineering Physics at Tsinghua University, Beijing (Wang et al., 2015). RMC has the capacity for lifecycle simulations of nuclear reactor cores. The BEAVRS benchmark was selected as an example and RMC was applied to a full core, two cycle burnup calculation of BEAVRS. All of the parameters given in the BEAVRS benchmark have been calculated and compared with the measured values of BEAVRS benchmark. For other parameters such as pin power distributions, they are compared to the results of other codes. The results of RMC agree well with the measured values of BEAVRS benchmark and also agree well with those of other codes. This was the first time for a Monte Carlo code to perform the full core, two cycle calculation of BEAVRS. This work paves the way for Monte Carlo codes in life cycle simulations of nuclear reactor cores.

Identification of penetration path and deposition distribution of radionuclides in houses by experiments and numerical model

In order to lift of an evacuation order in evacuation areas and return residents to their homes, human dose assessments are required. However, it is difficult to exactly assess indoor external dose rate because the indoor distribution and infiltration pathways of radionuclides are unclear. This paper describes indoor and outdoor dose rates measured in eight houses in the difficult-to-return area in Fukushima Prefecture and identifies the distribution and main infiltration pathway of radionuclides in houses. In addition, it describes dose rates calculated with a Monte Carlo photon transport code to aid a thorough understanding of the measurements. The measurements and calculations indicate that radionuclides mainly infiltrate through visible openings such as vents, windows, and doors, and then deposit near these visible openings; however, they hardly infiltrate through sockets and air conditioning outlets. The measurements on rough surfaces such as bookshelves implies that radionuclides discharged from the Fukushima-Daiichi nuclear power plant did not deposit locally on rough surfaces.

Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment

An MCNP–ORIGEN burn-up calculation code system, named MCORE (MCNP and ORIGEN burn-up Evaluation code), is developed in this work. MCORE makes use of the Monte Carlo neutron and photon transport code MCNP4C and nuclides depletion and decay calculation code ORIGEN2.1. MCNP and ORIGEN are coupled by data processing and linking subroutines. In MCORE, a so called “modified predictor corrector” approach is used. MCORE provides the capability of using different depletion calculation schemes and simulating fuel shuffling. Total nuclide density changes in active cells are considered in MCORE. The validity and applicability of the developed code are tested by investigating and predicting the neutronic and isotopic behavior of a “VVER-1000 LEU Assembly Computational Benchmark” at lattice level and a “Physics of Plutonium Recycling” fast reactor at core level (OECD-NEA).The comparison results show that the MCORE code predicts the nuclide composition within 5% accuracy and k∞ within 800pcm at the end of the burn-up for LEU assembly (40MWD/kg HM). For a fast reactor, the results obtained by MCORE are in the range of reported results except for 243Am. In general, MCORE results show a good agreement with the benchmark values.

MCNPX–BUCAL1 code to code verification through burnup analysis

The availability of accurate burnup data is an essential first step in any systematic approach to enhancement of economics, safety and performance of a research reactor. This first step requires the utilization of a well verified burnup code system. In this work a newly home-developed burnup code called BUCAL1 is presented. The code provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP (version 5c). BUCAL1 has the capability of using several depletion calculation schemes that do not exist in several other burnup code systems such as: shuffling, refueling and multicycles burnup calculation, in an automatic way.The accuracy and precision of BUCAL1 were tested for U-Zrh fuels, by a code to code verification with MCNPX2.7, by predicting the burnup parameters of the 2MW TRIGA Mark II Moroccan research reactor. Continuous energy cross section data from the more recent nuclear data evaluation ENDF/B-VII.0 as well as S(α, β) thermal neutron scattering functions distributed with the MCNP code were used. Analysis of the verification results shows that BUCAL1 is enough accurate to be used in burnup calculations.

三维中子-光子输运的蒙特卡罗程序MCMG

三维中子-光子输运的蒙特卡罗程序MCMG

Development of parallel monte carlo electron and photon transport (PMCEPT) code III: Applications to medical radiation physics

Minimizing the differences between dose distributions calculated at the treatment planning stage and those delivered to the patient is an essential requirement for successful radiotheraphy. Accurate calculation of dose distributions in the treatment planning process is important and can be done only by using a Monte Carlo calculation of particle transport. In this paper, we perform a further validation of our previously developed parallel Monte Carlo electron and photon transport (PMCEPT) code [Kum and Lee, J. Korean Phys. Soc. 47, 716 (2005) and Kim and Kum, J. Korean Phys. Soc. 49, 1640 (2006)] for applications to clinical radiation problems. A linear accelerator, Siemens’ Primus 6 MV, was modeled and commissioned. A thorough validation includes both small fields, closely related to the intensity modulated radiation treatment (IMRT), and large fields. Two-dimensional comparisons with film measurements were also performed. The PMCEPT results, in general, agreed well with the measured data within a maximum error of about 2%. However, considering the experimental errors, the PMCEPT results can provide the gold standard of dose distributions for radiotherapy. The computing time was also much faster, compared to that needed for experiments, although it is still a bottleneck for direct applications to the daily routine treatment planning procedure.

Validation of IRBURN calculation code system through burnup benchmark analysis

Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.

Semantic Web-based PMCEPT Monte Carlo Code Simulations in Medical Physics

PMCEPT is a three-dimensional parallel Monte Carlo electron and photon transport code based on the standard message passing interface. It has been intensively validated for use in diverse medical physics applications such as radiotherapy, dosimetry, nuclear medicine, X-ray diagnostics, and radiation protection calculations. Despite extensive validation, the Monte Carlo code may not be an easy tool to pick up for a user working in medical areas because it requires some advanced knowledge of information technologies. To overcome this diculty, we developed a function web based on recent semantic web technology. Recent developments of the semantic web make it possible to understand and satisfy various requests of people and machines to intelligently use web contents or even control other machines. This paper describes a specific functional web (including shell scripts) to run the PMCEPT code through a network and verifies the overall dosimetric accuracy of PMCEPT physics algorithms with a 6-MV beam model of the Siemens Primus linear accelerator. A medical professional with relatively little information technology experience can take advantage of the Monte Carlo simulation via the functional web running on our dedicated server with his or her PC (client). We compare calculations with measurements in water and poly methyl methacrylate (PMMA) phantoms for 6-MV Siemens Primus photon beams. The target geometry of a PMMA head and neck phantom was derived from the CT data (DICOM format). Maximum discrepancies were lesser than 2 % in the tail parts of the lateral dose distributions. This study also suggests that our functional web may have a wide range of scientific applications because it is able to control any computer code running on physically separate machines via the Internet.

Patient-Specific Monte Carlo Dose Calculations for High-Dose-Rate Endorectal Brachytherapy With Shielded Intracavitary Applicator

An integrated software platform was developed to perform a patient-specific dosimetric study on high-dose-rate (192)Ir endorectal brachytherapy. Monte Carlo techniques were used to examine the perturbation effects of an eight-channel intracavitary applicator with shielding and a liquid-inflatable balloon. Such effects are ignored in conventional treatment planning systems that assume water-equivalent geometries. A total of 40 Task Group 43-based rectal patient plans were calculated using the PTRAN_CT Monte Carlo photon transport code. The silicone applicator, tungsten or lead shielding, contrast solution-filled balloon, and patient anatomy were included in the simulations. The dose to water and dose to medium were scored separately. The effects of heterogeneities and uncertainties in source positioning were examined. A superposition calculation method using pregenerated Monte Carlo dose distributions about the shielded applicator in water was developed and validated for efficient treatment planning purposes. On average, metal shielding decreases the mean dose to the contralateral normal tissues by 24% and reduces the target volume covered by the prescribed dose from 97% to 94%. Tissue heterogeneities contribute to dose differences of <1% relative to the prescribed dose. The differences in the dose volume indices between dose to water and dose to medium-based calculations were <1% for soft tissues, <2% for bone marrow, and >20% for cortical bone. A longitudinal shift of +/-2.5 mm and a rotational shift of +/-15 degrees in applicator insertion reduced the target volume receiving the prescribed dose by </=4%. The shielded applicator improved dose conformity and normal tissue sparing; however, Task Group 43-based treatment planning might compromise target coverage by not accounting for shielding.

Calculation of the reactor neutron time of flight spectrum by convolution technique

It is a very complex and time-consuming process to simulate the nuclear reactor neutron spectrum from the reactor core to the export channel by applying a Monte Carlo program. This paper presents a new method to calculate the neutron spectrum by using the convolution technique which considers the channel transportation as a linear system and the transportation scattering as the response function. It also applies Monte Carlo Neutron and Photon Transport Code (MCNP) to simulate the response function numerically. With the application of convolution technique to calculate the spectrum distribution from the core to the channel, the process is then much more convenient only with the simple numerical integral numeration. This saves computer time and reduces some trouble in re-writing of the MCNP program.

TH-D-351-06: Comparison Between 2D Monte Carlo Modeled and Experimental Cone-Beam CT X-Ray Projections

Purpose: Fast and accurate modeling of cone-beam CT (CBCT) x-ray projection data can improve cone-beam CT (CBCT) image quality either by conditioning projection data prior to image reconstruction or by supporting rigorous comparative simulation studies of competing image reconstruction and processing algorithms. In this study, we compare Monte Carlo- computed x-ray projections with projections experimentally acquired from our Varian Trilogy CBCT imaging system for phantoms of known design. Method and Materials: Our recently developed Monte Carlo photon-transport code, PTRAN, was used to compute primary and scatter projections for cylindrical phantoms of known diameter (CatPhan and NA model 76-410) with and without bow-tie filter and antiscatter grid for both full- and half-fan geometries. The simulations were based upon measured 120 kVp spectra, beam profiles, and flat-panel detector (4030CB) point-spread functions. The beam-stop array method was used to acquire scatter and SPR distributions from the OBI images. The biasing of scatter measurements due to the long detector PSF tails was corrected either by a lead mask or by deconvolution. Computed projections were compared to flat- and dark-field corrected 4030CB images. Results: The simulated primary profiles agree with experiment within 3%, while the simulated scatter profiles agree within 8–10%. Both PSF measurements and mask measurements indicate that scatter radiation values can be biased by as much as 7% detector PSF tails. Conclusion: In agreement with the literature, the difference between simulated and measured projection data is of the order of 6–8%. Higher accuracy can be achieved mainly by improving the beam modeling and correcting the non linearities induced by the detector PSF. This project was supported in part by grants from Varian Medical Systems and NCI (R01 CA 75371).

Evaluation of 192Ir high-dose-rate rectal brachytherapy treatment plans using a fast Monte Carlo photon transport code

Purpose: To retrospectively evaluate the treatment plans of a rectal cancer patient cohort who received 192Ir high-dose-rate (HDR) brachytherapy with a metal-shielded intracavitary mould applicator. Using fast Monte Carlo (MC) techniques, dose perturbations due to applicator and anatomical heterogeneities and their effects on the target dose coverage were examined.

BrachyGUI: an adjunct to an accelerated Monte Carlo photon transport code for patient-specific brachytherapy dose calculations and analysis

A number of accelerated Monte Carlo (MC) codes have been developed in recent years for brachytherapy applications, one of which is PTRAN_CT. Developed as an extension to the well-benchmarked PTRAN code, PTRAN_CT can be used to perform efficient patient-specific dose calculations. The code can explicitly account for the patient geometry converted from computed-tomography (CT) images, as well as perturbations due to the brachytherapy applicator and seeds. We have developed a software tool called BrachyGUI that provides an integrated environment for preparing patient and treatment plan-specific input data files for PTRAN_CT. It also comes with dose calculation, analysis, and treatment planning capabilities. In this article, we will describe the interface of BrachyGUI with PTRAN_CT for CT-based calculations, and examine the calculation efficiency of PTRAN_CT. We conclude that it is now feasible to use PTRAN_CT for high dose rate brachytherapy treatment planning on a routine clinical basis.

Irradiation conditions of ADS beam window and implications for window material

Accelerator driven systems (ADS) are aimed at incineration of long living radioactive isotopes of spent nuclear reactor fuel, thus providing a solution for nuclear waste utilization. The idea is to couple a subcritical nuclear reactor with a high-energy proton accelerator. Protons interacting with the liquid metal target inside the reactor will produce neutrons with energy sufficient to transform long living fuel isotopes to those with shorter decay time. In some design variants a proton beam guide and window separate the vacuum of the accelerator from the liquid metal target. The window appears to be the most critical component of the whole facility as besides liquid metal corrosion it undergoes irradiation damage from incident protons, from protons and neutrons produced by spallation inside the target as well as from fission neutrons generated in the reactor fuel assemblies. In the present work we have evaluated irradiation conditions of the ADS beam window using Monte Carlo neutron, photon and charged particle transport code MCNPX. The code and the detailed geometric model of the experimental ADS (XADS) facility allow a realistic simulation of the spallation process and transport of generated nucleons as well as evaluation of various damage and operational characteristics like displacement damage, heat deposition, gas and spallation element production rates. Present results based on the reduced beam parameters provide a hope that apart from the liquid metal corrosion (not considered in the present paper) the window material could sustain full operation during the period of 3–4 month between replacements. These results may allow reconsideration of pro and contra of the window and the windowless XADS concepts.

SU‐FF‐T‐327: Monte Carlo Dose Calculations in Homogeneous and Heterogeneous Media: A Comparison Between the PMCEPT Code and the MCNP5, EGS4, DPM Codes and Measurements

The Monte Carlo method for high‐energy photon and charged particle transport is the most accurate means for predicting dose distributions in radiation treatment of patients. Owing to rapid development of computer hardware and network technologies the use of this method is not restricted only to big research centers any more. A new parallel Monte Carlo electron and photon transport (PMCEPT) code [Kum and Lee, J. Kor. Phys. Soc. 47 (2005) 716] for calculating electron and photon beam doses has been developed based on a three dimensional geometry defined by computerized tomography (CT) images and implemented on the Beowulf PC cluster. The PMCEPT code was validated on the homogeneous and multi‐layered targets for megavoltage electron beams by comparing with the results of experiments and calculations from conventional Monte Carlo codes of the MCNP5, EGS4, and DPM. The computing time of the PMCEPT code was approximately twenty times faster than that of MCNP5 on the IBM ThinkPad X40 (laptop) with 1.2‐GHz CPU and 512‐MB ram memory operated by RedHat Linux 9. The PMCEPT results in general agreed well with others in homogeneous and heterogeneous media within a maximum of 2–3 % error. At the conference, we will show various benchmark results for the PMCEPT. This work was supported, in part, by the SRC/ERC program of MOST/KOSEF (grant number: R11‐1999‐054).

SU‐FF‐T‐17: 3D Monte Carlo‐Based Treatment Planning Evaluation of Intracavitary HDR Balloon Catheter Brachytherapy

Purpose: To evaluate the dosimetric effects of missing tissue and tissue heterogeneities on intracavitary HDR balloon‐catheter brachytherapy using a novel 3D Monte Carlo‐based treatment planning study. Method and Materials: The PTRAN Monte Carlo (MC) photon transport code is used to generate 3D calculations of the doses arising from an HDR 192 Ir source in; Anatomy‐based MC simulations modeling the dosimetric effects of missing tissue beyond the skin surface and tissue heterogeneities (such as air pockets) adjacent to the balloon catheter (Mammosite); and in a large water phantom representing those calculations achievable with a conventional 3D brachytherapy RTPS. Each MC simulation employs a track length estimator scoring method in a 3D grid, with 2mm resolution in a 12cm×11cm×8cm volume, scoring 8×106 primary photon histories. The dose matrices from all MC simulation are registered (in the external beam module of a commercial RTPS) to a single, modeled patient CT, which is identical in location, scale, and shape to the patient volume described in the anatomy‐based MC simulations. Several target volumes (e.g. a 1cm margin on the Mammosite balloon) and a skin‐surface contour, are created. Dose‐volume histograms (DVH's) are computed for each volume, per each plan. Comparisons of the resultant DVH's quantify the dosimetric errors expected from routine clinical planning of Mammosite applications. Results: The missing tissue beyond the skin surface reduces target coverage by 1.5–3.5% and skin doses by 8–20%, when compared to calculations achievable with conventional 3D brachytherapy treatment planning systems. A 10cc airpocket neighboring the Mammosite balloon reduces doses to adjacent tissues &lt;0.5%. Conclusion: A detailed 3D Monte Carlo‐based brachytherapy RTP study of the dose prediction errors in balloon catheter HDR applications shows that the missing tissue beyond the skin surface reduces skin doses by as much as 20% and target coverage by as much as 3.5%.