Monte Carlo Transport Code Research Articles

Development of a hybrid parallelism Monte Carlo transport middleware on mesh geometry

Mesh-based Monte Carlo particle transport is suitable for complex geometry modeling and for multi-physics coupling calculations. Performing large-scale mesh-based Monte Carlo simulations requires a quite large amount of computation and memory space. Therefore, parallel Monte Carlo computing is crucial. In this study, a hybrid parallelism Monte Carlo transport middleware on mesh geometry was developed. The middleware can provide the following algorithms and functions: 1. hybrid parallelism of domain decomposition and particle parallelism; 2. particle data structure, which manages particle information and automatically performs asynchronous communication between decomposed domains; 3. subset data structure, which provides an efficient management strategy for materials and tally data; 4. particle tracking method on structured/unstructured mesh and k-dimensional tree (KD-Tree) acceleration algorithm; 5. condensed application programming interface for the rapid development of Monte Carlo code based on the middleware. We transplanted the fixed source module of the J Monte Carlo transport code (JMCT) which was based on constructive solid geometry (CSG) to the middleware for verification. The results indicated that the serial particle tracking efficiency of an unstructured mesh model was 6.56 times as high as that of MCNP6 and the serial simulation efficiency of the corresponding structured mesh model was basically equivalent to that of JMCT. The parallel efficiency of domain decomposition was higher than that of JMCT. We also calculated the shielding problem of the Qinshan No. 1 reactor through hybrid parallelism. The results were in good agreement with the experimental data, and the hybrid parallel efficiencies were higher than 65% on 4000 CPU cores.

Effect of ion-induced nuclear reactions on structure modification and radiolysis in LiF irradiated by 410 MeV 36S ions

The objective of this research is to study the structural damage in LiF crystals irradiated with 410 MeV 36S ions. We report on the appearance of a remarkable structural damage far (hundreds of μm) beyond the penetration depth (136 μm) of sulfur ions. Selective chemical etching on samples irradiated at room temperature revealed the formation of extended defects within the ion-irradiated zone and also beyond the range of sulfur ions. Based on AFM image analysis, the defects were identified as nano-size colloids (10 nm). In the irradiated zone, nanoindentation tests indicated a significant hardening effect. Annealing experiments showed manifestations of intense radiolysis both in the ion-irradiated and beyond the range areas with a maximum of around 620 K. Comparison of experimental results with those of performed model simulation using Monte-Carlo particle transport code FLUKA support the assumption that the cause of the observed long-range damage is irradiation by secondary particles created in inelastic nuclear collisions between incident sulfur ions and lithium and fluorine nuclei of the target atoms. Neutrons, γ-rays and protons are rated as the dominating components of the secondary particle flux. The results can be helpful for assessing the risks for damage delocalization and for reasonable selection of ions in the applications of ion technologies to optical materials like LiF.

Tritium Transport in WCLL Outboard Breeding Blanket of EU-DEMO Reactor Under Pulsed Plasma Operation

The accurate prediction of the tritium inventory and permeation fluxes in the water-cooled lithium-lead (WCLL) breeding blanket of the European DEMO fusion reactor is a key aspect for future thermonuclear power plant licensing. In this context, a tritium transport analysis is essential for the evaluation of the tritium retention in LiPb (15.7 at. % Li) and in the structural components and tritium permeation fluxes into the cooling water. This study presents a COMSOL Multiphysics model of a portion of the minimum functional unit of the outboard equatorial module of the WCLL. The neutronics analysis, performed with the MCNP Monte Carlo transport code, allows for the assessment of the tritium generation rate and the volumetric power deposition in the lithium-lead, both of which are used as an input for tritium transport modeling. Moreover, the buoyancy forces and the magnetohydrodynamic effect are also included. In order to take into account a pulsed operation for the DEMO reactor, a suitable algorithm was developed. The results show peak velocities in LiPb up to 24.1 mm with a maximum temperature at the first wall (FW) of 441°C. Most of the tritium is concentrated between the FW and the baffle plate, with partial pressures between 0.15 and 0.4 Pa considering Aiello’s correlation and in-house measurements of Sieverts’ constant on the HyPer-QuarCh II (HQ-II) device, whereas from 70 to 180 Pa adopting Reiter’s correlation. This has an impact also on the inventory in the Eurofer and in permeation fluxes to the water pipes. For Reiter’s correlation, the inventories in the Eurofer are about one order of magnitude lower with respect to LiPb, and the inventory in water is five orders of magnitude higher with respect to HQ-II and Aiello’s correlation.

High-fidelity neutronics and thermal-hydraulics analysis of helical cruciform fuel based on small modular FHR

Fluorine High-temperature Reactor (FHR) is a novel reactor type without water-cooling, but with a high outlet temperature. Small Modular FHR is convenient for modular assembly and transportation. Therefore, it is suitable for remote and water shortage areas to provide efficient and clean nuclear energy. The neutronics and heat transfer of the fuel are the key issues when the small modular FHR has high thermal power output. This research proposes the fuel – TRISO (TRi-ISOtropic) particles dispersed in Helical Cruciform Fuel (HCF). HCF can be self-supported in radial direction, no additional grid-spacers are required, which can decrease the pumping power and improve the efficiency of the reactor system. For small modular FHR, it can be disassembled, transported, and assembled better without grid-spacers and simplifying the core structure. The heat transfer capacity of the coolant also could be further improved because of the special structure of HCF. Based on the Monte-Carlo particle transport code MCNP, the neutronic characteristic from the assembly to the whole core of FHR is studied. By obtaining the fission power distribution in near-critical condition, the thermal-hydraulics of the single most heated HCF and 7 bundle HCF in the heated assembly are calculated. In order to preliminary analyze the thermal-hydraulics of the whole core scale, the thermal-hydraulics analysis of porous media is carried out for 1/6 core. The isotropic resistance tensor and thermal equilibrium porous model have been introduced with reasonable approximation, and porous resistance relation has been carried out. The radial power peak factor is 1.185 and the axial power peak factor is 1.435 in near-critical condition, which proves the arrangement of the HCF and the control rods can flatten the power distribution effectively. The average temperature of the coolant outlet in the active region of the core is 974.44 K, and the average temperature of the upper plenum outlet is 972.85 K. High coolant outlet temperature can ensure effective supply of high-temperature process heat. And the highest temperature of the fuel is 1271.49 K, which is lower than the limitation 1573 K. The parameters of this research can supply the neutronics and thermal-hydraulics references for further FHR design.

An internal coupling method between neutronics and thermal-hydraulics with RMC and CTF

To tackle the high-fidelity multi-physics coupling challenge in nuclear reactors, an internal coupling method was proposed and implemented to closely couple the neutron transport Monte Carlo code RMC and the sub-channel thermal-hydraulics solver COBRA-TF (CTF) using direct memory data transfers. In this coupling system, fixed-point Picard iteration scheme and Gauss-Seidel iteration method were implemented and evaluated. Two acceleration algorithms were proposed and both proved effective, including the hybrid MPI/OpenMP parallelism scheme and the asymptotic batch coupling method. The coupling system was validated against the VERA benchmarks and compared with literatures. The k eigenvalue differed by 7 pcm and the relative differences in the radial power profiles were less than 0.30%. Moreover, the simulation time was reduced by 48.1% due to quicker convergence with the two acceleration methods. Further analyses also proved the superior precision of this method, especially for asymmetric problems.

A simulation optimization design of the filter stack spectrometer for laser-plasma interaction experiment

Filter stack spectrometers are widely employed in laser facilities for the spectrum measurement of bremsstrahlung photons. However, this method suffers from large uncertainty of unfolding due to its intrinsic limit resolution. For this, an optimization study on filter stack spectrometer is conducted. This procedure is implemented by a hybrid particle swarm optimization and genetic algorithm (PSO-GA). Monte-Carlo particle transport code Fluka is used for the simulation of the response matrix. Gravel algorithm, based on the least-square method, is used for the unfolding. For mono-energetic photons, this optimized filter stack spectrometer design provides a better energy resolution. For continuous distribution, this optimized filter stack spectrometer design yields a narrower unfolding solution space in the presence of measurement error.

Development of transient Monte Carlo in a fissile system with β-delayed emission from individual precursors using modified open source code OpenMC(TD)

In deterministic and Monte Carlo transport codes, β-delayed emission is included using a group structure where all of the precursors are grouped together in 6 groups or families, but given the increase in computational power, nowadays there is no reason to keep this structure. Furthermore, there have been recent efforts to compile and evaluate all the available β-delayed neutron emission data and to measure new and improved data on individual precursors. In order to be able to perform a transient Monte Carlo simulation, data from individual precursors needs to be implemented in a transport code. This work is the first step towards the development of a tool to explore the effect of individual precursors in a fissile system. In concrete, individual precursor data is included by expanding the capabilities of the open source Monte Carlo code OpenMC. In the modified code – named Time Dependent OpenMC or OpenMC(TD)– time dependency related to β-delayed neutron emission was handled by using forced decay of precursors and combing of the particle population. The data for continuous energy neutron cross-sections was taken from JEFF-3.1.1 library. Regarding the data needed to include the individual precursors, cumulative yields were taken from JEFF-3.1.1 and delayed neutron emission probabilities and delayed neutron spectra were taken from ENDF-B/VIII.0. OpenMC(TD) was tested in a monoenergetic system, an energy dependent unmoderated system where the precursors were taken individually or in a group structure, and in a light-water moderated energy dependent system, using 6-groups, 50 and 40 individual precursors. Neutron flux as a function of time was obtained for each of the systems studied. These results show the potential of OpenMC(TD) as a tool to study the impact of individual precursor data on fissile systems, thus motivating further research to simulate more complex fissile systems.

A Mission to Mars: Prediction of GCR Doses and Comparison with Astronaut Dose Limits.

Long-term human space missions such as a future journey to Mars could be characterized by several hazards, among which radiation is one the highest-priority problems for astronaut health. In this work, exploiting a pre-existing interface between the BIANCA biophysical model and the FLUKA Monte Carlo transport code, a study was performed to calculate astronaut absorbed doses and equivalent doses following GCR exposure under different shielding conditions. More specifically, the interface with BIANCA allowed us to calculate both the RBE for cell survival, which is related to non-cancer effects, and that for chromosome aberrations, related to the induction of stochastic effects, including cancer. The results were then compared with cancer and non-cancer astronaut dose limits. Concerning the stochastic effects, the equivalent doses calculated by multiplying the absorbed dose by the RBE for chromosome aberrations ("high-dose method") were similar to those calculated using the Q-values recommended by ICRP. For a 650-day mission at solar minimum (representative of a possible Mars mission scenario), the obtained values are always lower than the career limit recommended by ICRP (1 Sv), but higher than the limit of 600 mSv recently adopted by NASA. The comparison with the JAXA limits is more complex, since they are age and sex dependent. Concerning the deterministic limits, even for a 650-day mission at solar minimum, the values obtained by multiplying the absorbed dose by the RBE for cell survival are largely below the limits established by the various space agencies. Following this work, BIANCA, interfaced with an MC transport code such as FLUKA, can now predict RBE values for cell death and chromosome aberrations following GCR exposure. More generally, both at solar minimum and at solar maximum, shielding of 10 g/cm2 Al seems to be a better choice than 20 g/cm2 for astronaut protection against GCR.

A new approach for uncertainty quantification in predictor-corrector quasi-static Monte Carlo transient simulation

Two different approaches are widely accepted for transient Monte Carlo (MC) simulation namely the Dynamic Monte Carlo (DMC) and the Predictor-Corrector Quasi-Static Monte Carlo (PCQS-MC). The MC transport code iMC developed at the Korea Advanced Institute of Science and Technology incorporates both approaches. In this paper, an original method for properly assessing the uncertainty of PCQS-MC is proposed and demonstrated using the iMC code. A detailed description of applying the PCQS method for transient MC calculation is presented with an emphasis on the origins of PCQS-MC uncertainties. The implementation of a quasi-static method, i.e., calculation of point-kinetics (PK) parameters, incurs an additional uncertainty for PCQS-MC calculation alongside the conventional stochastic MC uncertainty. Such quasi-static treatment-driven stochasticity cannot be recognized through the conventional MC sampling scheme, insinuating a significant underestimation of uncertainty without proper measures. To verify the findings, null-transient simulations have been performed for GODIVA and C5G7 benchmarks, where the former and latter problems require continuous and multi-group energy treatment respectively. For an improved uncertainty evaluation, a new sampling scheme referred to as the PK-sampling scheme is proposed where cycle-wise PK correction is made. Proper estimation of uncertainty can be made from the heap of cycle-wise corrected power through a unique screening process based on the null hypothesis testing. The proposed PK sampling scheme is tested for C5G7-TD transient benchmarks using iMC where the result plainly attests to the effectiveness of the new sampling approach.

Comprehensive Analysis of Streaming and Shutdown Dose Rate Experiments at JET with ORNL Fusion Neutronics Workflows

Current experimental fusion systems and conceptual designs of fusion pilot plants (FPPs) are growing in complexity and size. Several radiation metrics are crucial to the safe operation of fusion machines, including neutron flux streaming through openings and the shutdown dose rate (SDDR). Most current designs of advanced experimental fusion systems—and the most probable candidates for FPPs—are based on the tokamak concept, which is prone to neutron streaming through the myriad openings needed for diagnostic and support systems. SDDR is caused by decay gamma rays from radionuclides that become activated by neutrons during the operation of a fusion system that use deuterium-deuterium (DD), tritium-tritium, or deuterium-tritium plasma. Because computational tools have become essential for determining these radiation metrics, they must be validated against reliable and applicable experimental data. Experiments at the Joint European Torus (JET) provide a unique source of experimental data for validating computational tools and nuclear data used to determine SDDR and neutron fluxes in streaming-dominated geometries. This paper presents the comprehensive analysis of the high-performance DD JET SDDR, and streaming experiments performed using Oak Ridge National Laboratory (ORNL) fusion workflows. The computational results were compared with experimental results that consist of online SDDR measurements with ionization chambers and neutron fluence streaming measurements using thermoluminescent detectors. The ratio of calculated-to-experimental SDDR values ranges from 0.6 to 2.5, and the streaming results range from 0.5 to 8.0. Future work will include analyzing the JET 2021 DTE2 campaign alongside the integration of the Shift Monte Carlo transport code into all ORNL fusion neutronics workflows.

Pulsed-Neutron Die-Away Experiments for Plastics and Neutron Thermal Scattering Laws

Pulsed-neutron die-away (PNDA) experiments can be useful benchmarks to validate neutron thermal scattering laws (TSLs). The experiment uses a neutron generator to impinge a short (∼10−4 s), mono-energetic neutron pulse on a target sample. After the pulse, the neutron population within the sample moderates and reaches thermal equilibrium with a fundamental spatial mode and characteristic decay-time eigenvalue. The eigenvalue can be extracted from the experimental measurements of the neutron flux and used as an integral parameter in validation. For certain materials and geometric configurations, the eigenvalue is heavily influenced by thermal neutron scattering of only the target material. For that reason, a PNDA experiment can have a higher sensitivity to TSLs than is commonly available in critical experiments. Herein, we present results for a series of new PNDA experiments conducted at Lawrence Livermore National Laboratory with plastic materials, e.g. high-density polyethylene and Lucite. We compare the experimental integral parameters to simulated results and report trends in the biases. We evaluate the bias with different Monte Carlo transport codes (MCNP6.2 and Mercury) and show no significant differences between the results of the two codes.

Study on kinetic parameters of pebble bed reactor with TRISO duplex fuel

Thorium, in this case, 232Th has a higher thermal neutron capture cross-section than 238U, which means that more fertile isotopes can be transmuted and could lead to higher fissile isotope 233U. In addition, 233U has a good performance in the thermal spectrum. Theoretically, a nuclear reactor using thorium fuel can also last longer than one using uranium fuel. The use of TRISO duplex fuel is predicted to produce better neutronic behavior in a pebble bed reactor. This work aims to study the kinetic parameters of a pebble bed reactor with TRISO duplex fuel. The configuration of the TRISO duplex fuel pebble consists of an inner region filled with UO2 TRISO particles and an outer region filled with ThO2 TRISO particles surrounded by a graphite matrix of fuel pebble. Three configurations with volume fraction of UO2-ThO2 were considered in this study: 80-20 %, 75-25 %, and 70-30 %. The HTR-10 reactor was chosen as a reactor model because its geometry and material specifications are known. A series of calculations were conducted using the Monte Carlo transport code MCNP6 and ENDF/B-VII.1 nuclear data library. The calculation results were then analyzed to investigate the effect of UO2 and ThO2 compositions in TRISO duplex fuel on the kinetic parameters of the pebble bed reactor with various TRISO packing fractions of 1-50 %. It can be concluded that the utilization of TRISO duplex fuel in a pebble bed reactor could significantly affect the core multiplication factor and kinetic parameters caused by an increase in Th content. On the other hand, the TRISO packing fraction is taking part in neutron moderation since a lower packing fraction means higher moderation for fueled pebble.

FLUKAVAL – A validation framework for the FLUKA radiation transport Monte Carlo code

The FLUKA general purpose radiation transport Monte Carlo code being developed and maintained by CERN (https://fluka.cern) has adopted modern software development standards including a formal quality assurance process. This includes the FLUKAVAL testing framework that takes into account the specific needs of testing a Monte Carlo radiation transport simulation code. FLUKAVAL allows to submit, process and validate a large number of test cases for a new FLUKA version and compare the results against previous versions and reference data. It produces a quantitative and qualitative comparison and compiles a summary report over all selected test cases, which is reviewed before the new FLUKA version is released. The overall validation requires only few manual steps and makes use of large computing clusters to run simulations in parallel.

Measurement of delayed neutrons in a thermal nuclear reactor by means of a long run pile noise experiment in sub-critical state

The transfer function of a zero-power thermal reactor was successfully measured thanks to neutron noise techniques from 1 mHz to 160 Hz. During a month-long experimental campaign, the fluctuations of the neutron population in critical and subcritical configurations of the core were acquired using excore fission chambers and analysed through Cross-Power Spectral Density (CPSD) methodology. Firstly, the reactor’s kinetic parameters, i.e. prompt decay constant, effective delayed neutron fraction and generation time, were obtained at critical state. It required calibrating the reactor’s power, which was done by metal foil activation and measurement of the 235U fission rate. Secondly, these parameters were used to estimate the groups’ abundances of delayed neutrons from the CPSD measured in a sub-critical state. Fitting data with a point kinetic model was done with Bayesian inference - CONRAD and Stan programs were used. A very good agreement was found between experimental abundances and the ones computed with TRIPOLI-4 Monte-Carlo transport code and JEFF3.1.1 nuclear data library. Uncertainties on prior abundances between 6 % to 101 %, held mainly by nuclear data, were lowered down to 4 % to 54 %, depending on the delayed group.

Modelling of a SiC Based Detector for the Interpretation of 14.1 MeV Neutrons Measurements

Wide-bandgap semiconductor-based fast neutron detectors such as silicon carbide (SiC) seem to be a promising concept to meet the implementation requirements as well as the performance specifications for fusion and fission environments such as radiation hardness as well as thermal and mechanical stabilities. Beyond the problem of integration of the device in constrained environments, the issue of the quantitative response of the device in a mixed radiation field must be addressed. Therefore, the characterisation of the detector performances, according to the energy of the neutrons is a key first step needed for neutron detection and monitoring in mixed radiation environments with high levels of gammas and fast and thermal neutron fluxes emissions at large energy scales. This paper presents the interpretation of 14.1 MeV neutron measurements using a numerical model developed with the GEANT4 MonteCarlo transport code. The methodology used for a first attempt to estimate the fluence rate of incident neutrons is exposed and a calculation versus experiment ratio is determined.

Proposal of point-wise power reconstruction for IRT-4M fuel assemblies in coupled Monte-Carlo calculations

Current efforts to develop high-fidelity multi-physics tools for research reactors with IRT-4M fuel leads to the involvement of Monte-Carlo transport and subchannel thermo-hydraulic codes in the process of safety assessment. Monte-Carlo based neutronic codes are often used as a reference tool and are suitable for a prediction of a detailed power distribution. This paper proposes a method of the point-wise power reconstruction based on a Radial Basis Function (RBF) interpolation for complicated irregular geometries and compares different detector functionalities of the Serpent2 code from the time efficiency point of view. Furthermore, a preliminary validation of the applied method for IRT-4M in a subchannel code SUBSALS mesh is presented. This code is currently under the evelopment at the Research Centre Rež and the Czech Technical University in Prague and focuses on the thermohydraulic analysis of IRT type fuel assemblies.

Simulation and analysis of the secondary-particle-induced background of the follow-up x-ray telescope onboard Einstein Probe

The Einstein Probe (EP) is a small scientific satellite committed to time-domain astronomy and high-energy astrophysics. One of the main payloads in the EP is the Follow-up X-ray Telescope (FXT). To get a quantity of in-orbit data as a reference, data from the Insight-Hard X-ray Modulation Telescope (Insight-HXMT) were adopted. The Insight-HXMT is dedicated to perform hard x-ray all-sky surveys, which has a parallel in-orbit environment with the EP. In this paper, we simulated the in-orbit background of an FXT by the Particle and Heavy-Ion Transport code System, which is the first general-purpose heavy ion transport Monte Carlo code, which has a simulated range from 0.01 meV to 1 TeV. The influence from protons in the EP orbital environment and the secondary rays caused by the shielding materials around the detector have been mainly considered. The Super Monte Carlo Simulation Program for nuclear and radiation Processes has been used to define the materials of the shielding structure. An isotropic source of protons and a complete shielding structure have been constructed in this simulation. First, the spectrum of background between 0.5 and 15 keV has been calculated. Second, we simulated the types of reactions and the fluxes of secondary particles, and all of them were calculated with the same model in the first step. Finally, we considered the effectiveness of the outermost shielding with different materials. Among the calculated spectra of background, we found two conspicuous energy responses with high counts, separately located in 7.345 keV (5 eV) and 8.045 keV (5 eV). Compared with the in-orbit data from the low energy x-ray telescope in the Insight-HXMT, the simulated results are credible. Considering the combination of the simulated types of reactions and fluxes of secondary particles, we could mostly judge that the secondary photons brought the photoelectric effects, which caused the two conspicuous responses in the spectrum of background. In addition, copper was the most suitable material for the outermost shielding in this simulation.

Uncertainty Quantification of the Decay Gamma Source in Mesh-Based Shutdown Dose Rate Calculations

In fusion energy systems, part of the design effort is dedicated to the assessment of the shutdown dose rate (SDR) due to the decay photons that will be emitted from activated components. Monte Carlo transport codes are often used to obtain the neutron flux distribution in the problem domain. The neutron flux distribution is used in the rigorous 2-step (R2S) workflow to obtain the photon emission density distribution of decaying radionuclides. The photon emission density is then used as an input for a dedicated photon transport step to calculate the SDR. In this paper, the uncertainty of the decay gamma source due to the uncertainty of the neutron flux distribution in the R2S workflow is investigated. A scheme is developed to estimate the uncertainty of the decay gamma source, building on the concept of groupwise transmutation and using standard error propagation techniques. The applicability of the newly developed scheme is then demonstrated on one of the conceptual designs of the fusion nuclear science facility.

Numerical Simulation of the Intermediate-Time High-Altitude Electromagnetic Pulse

Research works on high-altitude electromagnetic pulse (HEMP) have attracted considerable attentions for its ability to disturb or damage the electronic and power systems. HEMP can be divided into the following three parts: early-time (E1), intermediate-time (E2), and late-time (E3). Most of the studies of HEMP are focused on E1, and few investigations have been done on E2 in the open literatures. However, the amplitude of E2 is about 10–100 V/m, and it is able to couple into the cables whose lengths are several kilometers. Therefore, E2 is also a threat to power systems. The simulation of E2, which can provide more accurate environmental parameters for research works on HEMP interactions with systems, is an important supplement to HEMP simulation. In this article, a numerical simulation method of E2 is presented which combines the Monte Carlo transport code and the 2-D EMP code to simulate both E1 and E2. In addition, the simulation results of E2 are given which are in qualitative agreement with the standard waveform. This work is helpful to fill the gap in the simulation of E2 at present and to simulate HEMP environment more accurately and completely.

OpenNeoMC: A framework for design optimization in particle transport simulations based on OpenMC and NEORL

We present a study for particle transport design optimization that combines OpenMC (Monte Carlo transport code) with NEORL (NeuroEvolution Optimization with Reinforcement Learning). The proposed open-source framework, OpenNeoMC, empowers particle transport with state-of-the-art optimization techniques to facilitate design optimization and efficient selection of the simulation settings. The new framework is designed to have great concision and usability in integrating OpenMC into a variety of optimization techniques such as differential evolution, evolution strategies, bat algorithm, ensemble of differential evolution variants, and others. We present two optimization applications, namely criticality search of small modular reactor (SMR) and maximum keff search of a nuclear assembly. The deviation from criticality found by OpenNeoMC can be as small as the order of 1 pcm in the criticality search of SMR. For the second problem, an optimal geometric configuration of the nuclear assembly is found efficiently. The results demonstrate that the framework can be a powerful open-source tool to solve large-scale optimization problems in expensive particle transport simulations, and scalable enough to other Monte Carlo codes.