Monte Carlo Neutron Transport Code Research Articles

Improving the Resolution and Computational Efficiency of Depletion Simulations on a TRISO-Fueled Microreactor

Modeling individual TRi-structural ISOtropic (TRISO) particles is possible using constructive solid geometry in Monte Carlo neutron transport codes (such as MCNP); however, the billions of surfaces required incurs considerable computational expense. Depletion simulations, in conjunction with Monte Carlo transport, are also inherently slow, requiring many individual criticality calculations to approximate time-dependent behavior. Additionally, choosing acceptable time step sizes, spatial zoning, and other depletion parameters involves prior knowledge and/or iteration by the user. The slow and complex nature of depletion calculations greatly increases the cost and duration of the design iteration process. The importance of tracked isotopes, spatial zones, and time stepping during depletion simulations is evaluated globally using neutron multiplication factor, total isotope inventory, and burnup. Different depletion schemes are compared using the TRISO-fueled Snowflake microreactor. Woodcock Delta Tracking and a Reactivity-equivalent Physical Transformation (RPT) are investigated as potential approaches to increase the speed of Monte Carlo neutron transport calculations involving TRISO fuel. The Monte Carlo N-Particle® Code (MCNP®) Version 6.3 is used to obtain the transport solution, and depletion calculations are performed using its internally coupled version of CINDER90. A prototype MCNP delta tracking module under development at Los Alamos National Laboratory is also used. Although a single spatial depletion region is sometimes acceptable, large errors in core lifetime and total 238Pu, 241Pu, and 242Pu mass are possible. The prototype delta tracking module significantly speeds up criticality calculations; however, depletion calculations are not always faster than those performed with surface tracking. The RPT method is effective at reducing CPU times in both criticality and depletion calculations, without significantly impacting neutron multiplication factor and total isotope inventory. The necessary number of time steps to converge total isotope inventory was also explored; the required number and spacing varied greatly depending on the objective of the depletion simulation.

Possible criticality scenario and its mechanism of the Windscale Works criticality accident in 1970 analyzed by computational fluid dynamics and Monte Carlo neutron transport

The Windscale Works criticality accident in 1970 resulted from mixing an aqueous solution with an organic solvent with different plutonium concentrations and densities. Although this accident has been studied using improved computer capabilities in recent years, a precise criticality scenario has not yet been identified. This study aims to clarify a possible criticality scenario of the accident—the time variation of reactivity and its mechanism. The accident was simulated by combining the multiphase computational fluid dynamics solver of OpenFOAM and the delta-tracking-based Monte Carlo neutron transport code Serpent2. Consequently, the periodic uneven arrangement of fluids might have caused oscillations in neutron leakage and absorption, resulting in periodic wavy reactivity changes. Furthermore, the emulsion, which was thought to be the primary cause, might not be the dominant mechanism for reactivity change, although it contributed to the criticality of the accident.

Microstructure, mechanical and corrosion property of neutron and γ-ray shielding (Gd2O3+W)/Al composites

In this study, a neutron and γ-ray absorbing (Gd2O3+W)/Al composite was fabricated by vacuum sintering. Firstly, through Monte Carlo Neutron and Photo Transport Code (MCNP) calculations, the content of Gd2O3 and W in the composite was determined to be 10 wt% and 25 wt%, respectively. Then, the influence of sintering temperature and sintering time on the microstructure, density, compression properties and electrochemical corrosion behavior of the (Gd2O3+W)/Al composites were investigated. The results indicate that sintering temperatures below 600 °C and sintering times shorter than 2 h led to insufficient density of the composite. Conversely, sintering temperatures above 600 °C and sintering times longer than 2 h caused W to agglomerate, resulting in a decrease in the density of the composite and a deterioration of its electrochemical performance. Therefore, the optimal sintering temperature and sintering time for the (Gd2O3+W)/Al composites were determined to be 600 °C and 2 h, respectively. Finally, the immersion corrosion performance in H3BO3 solution of the composite fabricated by the optimal process parameters was studied in detail. It was found that pitting was the main corrosion form of the (Gd2O3+W)/Al composite, and an oxidation film mainly composed of Al2O3 was formed on the surface.

Utilization of Magnesium Chloride (Mg-Cl) cycle for Hydrogen Production of SOMBRERO Fusion Reactor

The aim of this study is to investigate the hydrogen production potential of the Mg-Cl (magnesium chloride) cycle in the SOMBRERO fusion reactor. Three different percentages of UO2 nuclear fuel, namely 2%, 6% and 10%, have been used while keeping the fuel zone thickness constant. The performance of the SOMBRERO fusion reactor has been considered statically. The neutronic calculations have performed with Monte Carlo neutron transport code. Firstly, it has been determined that the tritium breeding ratio (TBR) and energy multiplication factor (M) values depends on neutronically. Secondly, the amount of hydrogen production with by using energy multiplication factor (M) have been performed. The highest hydrogen production has been obtained as 8,12687 kg/s for 10% UO2 fuel ratio.

Development of Two-Step Method for Fast Chloride MSR Neutronics

This work presents neutronics models of a small and large fast-spectrum molten chloride-salt reactor. The models are similar to designs being pursued by industry, and they may serve as generic preconceptual and simplified neutronics models that provide information for decision making in licensing-related areas. The two models were created using Serpent, a Monte Carlo neutron transport code, and Moltres, a neutron diffusion core simulator tool. Specifically, this study focused on exploring the applicability of diffusion theory to fast molten salt reactor (MSR) models, the capabilities of an open-source, MSR-oriented simulation tool (Moltres), and optimal energy-group structures. The proposed two-step method involves group-constant generation with Serpent and a multigroup diffusion solution by Moltres. Three energy-group structures were applied. The accuracy of the solutions was determined through comparisons between the two-step and Monte Carlo flux and multiplication factor solutions. The findings indicated diffusion theory captures neutronics with minimal error for the large MSR and yielded best results with the 27-group structure. The 27-group structure yielded an average group flux error below 2% and keff agreement between diffusion and transport solutions within 30 pcm. The accuracy of the two-step method decreased for the very small (high-leakage) fast chloride MSR, but the neutronics were captured acceptably well with the 33-group structure. In addition to exploring the capabilities of Moltres, this work contributes to the sparse literature involving open-source models of fast-spectrum MSRs. Future work is noted as expanding the capabilities of the neutronics models to incorporate thermal hydraulics.

Benchmarking exercise of indigenous Monte Carlo neutron transport code PATMOC for hexagonal full core reactor geometry

An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The code has been extensively tested for its algorithmic accuracy against numerous international numerical benchmarks, has also been validated on practical nuclear reactor systems like Advanced Heavy Water Reactor Critical Facility (AHWR-CF) and Apsara-U pool type research reactor located at BARC, and is now in regular use for day-to-day reactor physics studies. These reactor systems and most of the benchmark problems used for the verifications and validation of the code have square lattice arrangement in the core. In this paper, we present the results of verification of PATMOC code on whole-core reactor system with hexagonal lattice arrangement of fuel elements. In this exercise, we have analyzed the High Temperature engineering Test Reactor (HTTR) whole-core benchmark for integral and differential parameters at both lattice and core level. The results demonstrate the algorithmic accuracy of the code on hexagonal reactor systems also. This exercise will help make use of this code for reactor physics design studies also for the reactor systems with hexagonal arrangement.

Analysis of Kobayashi benchmark with indigenous Monte Carlo neutron transport code PATMOC

The solution of the neutron transport equation is the basic input for the reactor physics design of a nuclear reactor system. Because of the complexities of geometry and cross-section data, the neutron transport equation is generally solved using numerical methods. One of the difficulties in the solution using these methods concerns the accuracy of distribution of neutron flux in the system containing void regions in a highly absorbing medium. The difficulty arises because of the issue in the continuity of flux at the void and material interface. There is a sudden and large change in the flux at this interface. Kobayashi benchmarks are widely used problems for testing the ability and accuracy of reactor physics codes to compute neutron flux distribution on such systems. An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The Kobayashi benchmarks have been used to test the PATMOC code to verify its accuracy in flux computation for systems with voids in-between high absorbing materials. Our results show that the PATMOC estimated values of flux distribution in the system compare very well with the reference results provided with the benchmark. In this paper we present the detailed results and analyses of this benchmark with PATMOC code.

Neutronic and thermal-hydraulic calculations under steady state and transient conditions for a generic pebble bed fluoride salt cooled high-temperature reactor

Here, we develop multiphysics coupled neutronics/thermal-hydraulics models for a recently available generic Fluoride-salt-cooled High-temperature Reactor (FHR) at equilibrium core burnup conditions with reactivity control mechanisms. We utilized the pebble bed reactor transient tool KP-AGREE with multigroup homogenized cross-sections generated by the continuous energy Monte Carlo neutron transport code Serpent to simulate transient events for an FHR. Comparing global reactor parameters and safety characteristics between Serpent and published results to standalone neutronic calculations in KP-AGREE verifies our initial conditions for our transient models. This work explores three separate transients: a loss of flow accident, an overcooling accident, and a total control rod withdrawal accident. The boundary conditions of these scenarios come from either transient benchmark definitions of a gas-cooled pebble bed reactor or transient analysis of a similar FHR design. For the loss of flow accident, our predicted peak fuel and coolant outlet temperatures show a similar response to published values, albeit slightly higher, with differences between values attributed to our modeling assumptions. The overcooling transient simulation predicts a 52 K temperature increase in the maximum fuel kernel temperature and an approximately 32.6% increase in total core power due to the negative coolant temperature feedback of the FHR design. Our control rod removal transients predicted a 235 K increase in maximum fuel kernel temperature and a 107.5% increase in total core power. For all transient cases, the fuel temperature remained significantly lower than the temperature limit, but coolant outlet temperatures for the reactivity-initiated accident exceeded external core material temperature limits.

OpenMC interpretation and analysis of two series highly-enriched-uranium thermal-spectrum benchmarks for nuclear data validation

This study aims to provide additional validation of nuclear data performance through a comparative analysis of the evaluated nuclear libraries ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.3, and JEFF-4 T1. The study utilizes the OpenMC Monte Carlo neutron transport code and integral benchmarks provided by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) which have not been interpreted since their first modeling. The criticality benchmarks assemblies used in this study cover 2 series which are fueled with Highly Enriched Uranium (HEU) in Metal form (MET) and have Thermal neutron spectra (THERM), referred to as HMT-026 and HMT-027, respectively. The results show that for the HMT-026 benchmarks, ENDF/B-VII.1 and ENDF/B-VIII.0 have results close to experiments. The best global agreement with the HMT-027 experiments was found to be ENDF/B-VII.1. Through relevant critical and nuclear data sensitivity analysis, it is found that Be and BeO are respectively of crucial importance for HMT-026 and HMT-027 and the nuclear data of 9Be (especially for ENDF/B-VIII.0) should be reviewed. Additionally, it is worth noting that the Thermal Scattering Law (TSL) of BeO has a great influence on the keff calculation for the HMT-027 benchmarks, while the influence of TSL of metal Be is negligible for the HMT-026 experimental benchmarks.

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.

Improvement of Geant4 Neutron-HP package: Doppler broadening of the neutron elastic scattering kernel

Whether it is for shielding applications or for safety criticality studies, numerically solving the neutron transport equation with a good accuracy requires to precisely estimate the Doppler broadened elastic scattering kernel in the thermal and epithermal energy range of neutrons travelling in a free gas. In Geant4, low energy neutrons are transported using evaluated data libraries handled by the Neutron High-Precision (Neutron-HP) package. Version 11.00.p03 of the code features in particular the Doppler broadened elastic scattering kernel, provided by the so-called ‘Sampling of the Velocity of the Target’ (SVT) method. However this latter fails for resonant heavy nuclei such as 238U and can severely impact the solving of the Boltzmann equation in fissile media. To overcome this shortcoming, the Doppler Broadening Rejection Correction (DBRC) method has been implemented in Geant4 and successfully validated with the reference Monte Carlo neutron transport code Tripoli-4® (version 11). This development will be taken into account in the next release of the code. All the validations have been performed with an automated benchmark tool which has been designed to support the quality assurance of the Geant4 Neutron-HP package. This tool is currently available on an ad hoc Gitlab repository and will be included in Geant4.

Preliminary neutronics and thermal analysis of a heat pipe cooled traveling wave reactor

Heat pipe cooled traveling wave reactor (HPTWR) is a novel concept of heat pipe reactor, which adopts neutron breeding wave to pursue Uranium-Plutonium breeding. In this paper, the cladding material, power output, and axial breeding fuel region length of the HPTWR are optimized in order to achieve the safe operation of reactor with reactor neutronics and HP thermal analysis. Monte Carlo neutron transport code RMC is used to obtain the neutronics results (reactivity, consumption and breeding of nuclear fuel, and power distribution). A heat pipe (HP) code based on thermal resistance–capacitance network model (TRCNM) is developed to analyze the heat transfer characteristics of HP using the power distribution of fuel element obtained by RMC code in the optimized HPTWR. Results show the optimized HPTWR can achieve 70 MWth power output and 46 years continuous operation. The highest temperature peak of HP wall in the highest power fuel element of reactor appears at beginning of cycle (BOC) and is about 1800 K. Due to the flattening of axial power distribution with propagation of traveling wave, the temperature peak of HP wall gradually decreases to about 1784.6 K at end of cycle (EOC), which can reduce the thermal stress of HP wall and improve the safety of reactor. The corresponding heat transfer margin of HP in the highest power fuel element is always above 19.0% during the whole reactor operation, which helps to ensure the safe operation of the optimized HPTWR.

Validation of the burnup code MOTIVE with respect to fuel assembly decay heat data

The burn-up code MOTIVE is a 3D code for fuel assembly inventory determination developed at GRS in recent years. It modularly couples an external Monte Carlo neutron transport code to the in-house inventory code VENTINA. In the present publication, we report on the validation of MOTIVE with respect to full-assembly decay heat measurements of light water reactor fuel. For this purpose, measurements on pressurized water reactor and boiling water reactor fuel assemblies from different facilities have been analyzed with MOTIVE. The calculated decay heat values are compared to the measured data in terms of absolute and relative deviations. These results are discussed and compared to other published validation analyses. Moreover, the observed deviations between measurements and calculations are analyzed further by taking into account the results of the validation of nuclide inventory determination with MOTIVE. The influence of possible biases of calculated nuclide densities important to decay heat at the given decay times are investigated and discussed.

TOUCANS: A versatile Monte Carlo neutron transport code based on GEANT4

This paper reports the development of TOUCANS, a new Monte Carlo neutron transport code fully written using the GEANT4 toolkit. It aims at modeling complex systems easily and rapidly, thanks to a simple key–value input file. While its main focus is the transport of low energy neutrons – below 20 MeV – using evaluated cross-section libraries, it also takes advantage of the state-of-the-art capabilities offered by GEANT4 regarding the modeled cross-sections for practically all particles at all energies. Hence it gives access to non-specialists to a wide variety of applications involving different particles, such as the simulation of compact accelerator based neutron sources (CANS) which initially motivated its development. It leverages on recent implementations performed within GEANT4 to provide both shielding calculation capabilities (using Adaptive Multilevel Splitting, which is a full y automated variance reduction technique) and criticality calculation capabilities (to estimate the keff of a multiplicative system). It also benefits from a tight coupling with CADMesh to readily import CAD models, with NCrystal to extend the support of thermal neutron transport to crystals, with FIFRELIN to increase compound nucleus de-excitation accuracy, and with FUNZ to provide to the user parametric simulation capabilities (such as multi-objective algorithms and calculation of Pareto sets). Thanks to recent improvements of the neutron physics within the GEANT4 Neutron-HP package, the TOUCANS code can be considered on-par with industrial neutron transport codes such as Tripoli-4® or MCNP in term of neutron physics accuracy. Its sources should be released soon amongst GEANT4 advanced examples.

Development of multiphysics coupling system for nuclear fuel rod with COMSOL and RMC

Coupling fuel performance and neutronics can help improve the prediction accuracy of fuel rod behavior, which is important for fuel design and performance evaluation. A fuel rod multiphysics coupled system was developed with multiphysics software COMSOL and 3D Monte Carlo neutron transport code RMC. The fuel performance analysis module was built on top of COMSOL with the ability to simulate the fuel behavior in two-dimensional axisymmetric (2D-RZ) or three-dimensional (3D) mode. RMC was innovatively wrapped as a component of COMSOL to communicate with the fuel performance analysis module. The data transferring and the coupling process was maintained using COMSOL’s functionality. Two-way coupling was achieved by mapping power distribution and fast neutron flux fields from RMC to COMSOL and the temperature and coolant density fields from COMSOL to RMC. A fuel rod pin lattice was modeled to demonstrate the coupling. Results show that the calculated power and temperature distributions are reasonable. Considering the flexibility of the coupled system, it can be applied to the performance evaluation of new fuel design.

Determination of neutron flux redistribution factors for a typical pressurized water reactor ex-core measurements using Monte Carlo technique

In a typical pressurized water reactor, neutron detectors located outside the reactor core monitor reactor power. In addition, they are also used to measure the reactivity of the control rods. A novel approach to calculate the ex-core neutron detector response in a typical pressurized water reactor using the Monte Carlo technique is presented. A detailed ex-core model of the Krško nuclear power plant was developed using the Monte Carlo neutron transport code MCNP. Due to the location of the ex-core neutron detectors, the hybrid code ADVANTG is used to generate variance reduction parameters to accelerte the convergence of the results outside the reactor core. To use ADVANTG, the fixed neutron source had to be reconstructed from the criticality core calculation. This paper presents the sensitivity analysis of the response of the ex-core detectors to the neutron data libraries used, the description of the fixed neutron source and the ADVANTG parameters. It was found that a pin-wise description of the neutron source for at least two rows of fuel assemblies at the core periphery is necessary for accurate results. Our results show the importance of a correct description of the prompt neutron spectra in the high energy region and the impact this has on the response of the ex-core detectors. The method in which the prompt neutron fission spectra for important fission nuclides are weighted by the calculated reaction rates has been shown to be the best approximation, with deviations from the reference calculation within statistical uncertainty. The effect of nuclear data libraries on the response of the ex-core detector was investigated, and the difference between the ENDF/B-VII.0 and the ENDF/B-VIII.0 nuclear data libraries was ∼11%. When the deficient evaluation of the 56Fe isotope included in the ENDF/B-VIII.0 nuclear data library was replaced by the improved evaluation from the IAEA INDEN project, the differences decreased to ∼3.7%. In addition, neutron flux redistributions due to control rod movement were investigated and flux redistribution factors were updated using Monte Carlo particle transport methods. The reaction rate redistribution factors obtained with methods presented in this paper are within 1% agreement with the currently used factors.

A detection system for accurate ([formula omitted], n) neutron counting measurements of low-rate ([formula omitted], n) neutron sources

The determination of the (α, n) neutron yield of (α, n) neutron sources with low absolute total neutron emission rates is a challenging task given the complexity of the (α, n) neutron production mechanism, experimental limitations involved with measurements of weak neutron sources, and, for sources containing special nuclear material, concurrence of the production of (α, n) and fission neutrons. However, (α, n) neutron sources are prevalent in a variety of nuclear engineering disciplines – such as nuclear fuel cycle engineering, nuclear safeguards and nonproliferation, and nuclear waste management – and as such, characterization of physical (α, n) neutron sources and accurate representation of (α, n) neutron production in neutron transport codes are critical aspects of these disciplines. In this work, a moderated 3He neutron detector array with high fidelity listmode data acquisition capability was developed for measurements of low-rate (α, n) neutron sources to support (α, n) reaction nuclear data validation needs for nonproliferation applications as well as to provide experimental data for benchmarking (α, n) neutron source modeling methodologies in Monte Carlo neutron transport codes. Measurements were performed of a calibrated Eckert & Ziegler AmBe (α, n) neutron source individually and simultaneously with 252Cf sources of different strengths to simulate pure (α, n) and mixed (α, n) + fission source conditions. Total neutron counting and coincidence analysis methods developed for system data analysis were applied to the measured data, and deduced (α, n) neutron yield values agreed to within 1% of the with the vendor-calibrated AmBe source (α, n) neutron yield traceable to international standards for all measurements. Total (α, n) neutron yield uncertainties for source conditions dominated by (α, n) neutron emission did not exceed 3%.

Development and preliminary verification of a Monte Carlo neutron transport code IMPC-Neutron

A Monte Carlo neutron transport code IMPC-Neutron has been developed for Accelerator Driven Advanced Nuclear Energy System (ADANES) in Chinese Academy of Sciences. Based on the geometry module and initialization module provided by open source Monte Carlo code OpenMC, IMPC-Neutron adopts Unionized Grid method and Delta-tracking method to improve the speedof simulation. Besides, in order to reduce the memory consumption,a low-memory mode is developed in IMPC-Neutron to satisfy different computing conditions. A critical benchmark of ICSBEP and three full reactor core model benchmarks are used to verify the reliability andacceleration effect of IMPC-Neutron. The results show that the calculation results of IMPC-Neutron are in good agreement with OpenMC, and the acceleration efficiency is higher.

Investigation on the Use of the Monte Carlo Iterative k-Source Scheme for the Study of Neutron Subcritical Multiplication

Monte Carlo neutron transport codes have traditionally used a fixed-source scheme to simulate a subcritical system with an external source. The efficiency of this scheme is known to depend on the subcriticality level: The lower the subcriticality is, the worse is the efficiency. We have investigated an alternate iterative scheme, namely, the Monte Carlo iterative k-source (IKS) scheme, for the study of neutron subcritical multiplication. Our results show that the iterative scheme not only is as accurate, effective, and computationally efficient as the fixed-source scheme but also has the additional advantage of being weakly dependent on the subcriticality level. Also, the efficiency of this scheme is unaffected by the change in the location of the external source, unlike the fixed-source scheme where the efficiency decreases as the source is moved away from the fissile core center. The algorithm of this scheme is very similar to the algorithm of the eigenmode iterative scheme and hence can be easily implemented in the existing Monte Carlo codes. Our work establishes the validity and accuracy of the Monte Carlo IKS scheme, and with its incorporation in the production-level codes, it can be used for the physics design and analysis of accelerator-driven subcritical systems.

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.