Search Results (24)

A design for a small chloride salt fast reactor (sm-MCFR) is presented through the integration of molten salt reactor and small reactor technologies, targeting efficient transmutation of transuranic (TRU) elements in spent nuclear fuel and rapid reactor deployment. The feasibility exploration and research on the design boundaries of sm-MCFR will be conducted in this article. The core adopts a dual-fluid configuration, in which the fuel salt and coolant circulate independently. Chloride salt is selected as the fuel carrier due to its high solubility for heavy metal nuclides and the low neutron absorption cross-section of chlorine, which help to form a hard fast-neutron spectrum and thereby enhance transmutation efficiency. The cooling system employs a direct supercritical carbon dioxide (s-CO₂) cycle, simplifying the overall layout. For the neutronics design, simulations were carried out using the TMCBurnup (TRITON MODEC Coupled Burnup Code). By adjusting the core geometry, fuel salt composition, and reprocessing strategy, the sm-MCFR achieves a hard fast-neutron spectrum but also demonstrates good potential for fuel utilization. In terms of thermal–hydraulic design, the heat exchange effect of the reactor core can be improved by adjusting the proportion of the coolant and the flow direction. The sm-MCFR is expected to become a promising candidate for advanced small reactors that have potential applications in nuclear waste transmutation and distributed energy generation. Full article

Transient analyses of liquid-fueled molten salt reactors involve strong coupling among neutronics, delayed neutron precursor transport, thermal–hydraulics, and solid heat transfer, leading to high computational costs for repeated high-fidelity simulations. To enable fast multi-physics prediction at unseen operating conditions, a parametric non-intrusive reduced-order model (ROM) combining proper orthogonal decomposition (POD) and higher-order dynamic mode decomposition (HODMD) is developed. Coupled full-order snapshots generated from an OpenFOAM-based one-eighth symmetric core model based on a simplified MSRE benchmark configuration are used to construct reduced representations for 11 physical fields. The POD truncation rank, HODMD delay dimension, and interpolation model are selected using leave-one-out cross-validation, with polynomial, radial basis function, and Gaussian process regression models considered as interpolation candidates. For unseen parameter points, the model maintains high accuracy in both the interpolation stage and the temporal extrapolation stage. In the temporal extrapolation stage, the highest mean relative $L_2$ error for the inlet-temperature-step case is $2.112\%$, whereas all mean relative $L_2$ errors for the inlet-velocity-step case remain below $0.177\%$. The results indicate that, under the present cases and parameter settings, the proposed framework provides an accurate and rapid surrogate for multi-physics transient prediction. Full article

Molten salts act as fuel carriers and coolants in liquid-fueled molten salt reactors (MSRs), characterized by strong coupling between neutronics and thermal hydraulics (N-TH) in practical MSR operations. In this study, an in-house light water reactor static and transient analysis code, RESTA-3D, has been extended and applied to MSR transient safety analysis. A parallel multi-channel TH model and a neutron kinetics model incorporating the transport of delayed neutron precursors were implemented into RESTA-3D to account for the MSR-specific N-TH coupling characteristics. Few-group cross-section parameters were generated by the TMSR-LINK code and tabulated for use in RESTA-3D to support MSR transient analysis. The code system was verified against simulation results from well-established MSR dynamics codes and validated against experimental data from the MSRE (Molten Salt Reactor Experiment), covering steady-state temperature distributions, fuel pump-driven transients, and the MSRE natural convection test. Good agreement of the improved RESTA-3D results with the experiment data of MSRE was confirmed, with key parameters such as temperature within a 1% deviation margin, thereby confirming that RESTA-3D is suitable for MSR dynamics analysis. Furthermore, this code was applied to assess the transient characteristics of a 2 MWth thorium-based molten salt reactor (TMSR). The core characteristics, including the inlet fuel overcooling and overheating, unprotected fuel pump start-up and coast-down, were simulated and discussed, indicating that the 2 MWth TMSR design possesses high inherent safety. Full article

Compared to conventional solid-fueled reactors, the liquid fuel transport in molten salt reactors (MSRs) leads to a strong coupling between thermal-hydraulics and neutronics. To enable system-level analysis of MSR, this study focuses on the main loop of the Molten Salt Reactor Experiment (MSRE). A system model is developed using the open-source, multiphysics modeling platform Modelica/TRANSFORM. The model is validated against ORNL experimental data under various conditions, including zero-power pump start/stop, natural circulation. In addition, the xenon transport behavior is compared with predictions from a two-region analytical model. Results indicate that the number of discretized core nodes significantly influences the estimation of delayed neutron precursor (DNP) losses due to fuel circulation. The applicability of the ANSI/ANS-5.1 decay heat model, originally developed for light water reactors, is confirmed to be conservative when applied to MSRE conditions. Finally, natural circulation behavior with decay heat transport is further analyzed. Full article

As the global energy demand continues to grow and the pursuit of clean and sustainable resources intensifies, nuclear energy stands out as a secure, reliable, and low-emission solution. The complexity of nuclear power plant behavior under various operating conditions necessitates advanced simulation tools capable of capturing the interplay between multiple physical phenomena. Among these, multi-physics coupling, particularly between neutronics and thermal hydraulics, is a well-established approach for accurately modeling transient scenarios with strong feedback effects. In this context, PARCS and TRACE codes, developed by the U.S. Nuclear Regulatory Commission, are widely used for coupled neutronic/thermal-hydraulic analyses and can be operated via the SNAP graphical interface. However, the current version of SNAP does not support automatic coupling for hexagonal core geometries, such as those found in VVER-type reactors. To address this limitation, a dedicated tool was developed to facilitate the coupling process between PARCS and TRACE for hexagonal cores. The proposed methodology was tested through the simulation of a rod ejection accident in a VVER-1000 reactor, demonstrating the validity of the methodology and confirming that the multi-physics approach provides more accurate, best-estimate results. Full article

This work presents a high-fidelity simulation of the Peach Bottom turbine trip (PBTT) benchmark using the Virtual Environment for Reactor Applications (VERA), a multiphysics reactor modeling tool developed by the U.S. Department of Energy’s Consortium for Advanced Simulation of Light Water Reactors energy innovation hub. The PBTT benchmark, based on a 1977 transient event at the end of cycle 2 in a General Electric Type-4 boiling water reactor (BWR), is a critical test case for validating core physics models with thermal feedback during rapid reactivity events. VERA was employed to perform end-to-end, pin-resolved simulations from conditions at the beginning of cycle 1 through the turbine-trip transient, incorporating detailed neutron transport, fuel depletion, and subchannel thermal hydraulics. The simulation reproduced key benchmark observables with high accuracy: the peak power excursion occurred at 0.75 s, matching the scram time and closely aligning with the benchmark average of 0.742 s; the simulated maximum power spike was approximately 7600 MW, which is within 3% of the benchmark average of 7400 MW; and void-collapse dynamics were consistent with benchmark expectations. Reactivity predictions during cycles 1 and 2 remained within 1500 pcm and 400 pcm of criticality, respectively. These results confirm VERA’s ability to model complex coupled neutronic and thermal hydraulic behavior in a BWR turbine-trip transient, which will support its use in future studies of modeling dryout, fuel performance, and uncertainty quantification for transients of this type. Full article

Most Nuclear Power Plants (NPPs) are designed for baseload operations, maintaining a steady power output at 100%, except during planned maintenance and refueling. However, in countries like France, Slovakia, and Korea, where nuclear power is a major source of electricity, integrating nuclear energy with intermittent renewables is crucial for stable power generation. This integration necessitates daily power adjustments by NPPs in response to grid demands, a process known as a Load Follow Operation (LFO). Such a process introduces strong interdependencies between thermal–hydraulic and neutron–kinetic parameters, coupled with the three-dimensional movement of Control Element Assemblies (CEAs) and Xenon dynamics, which pose safety challenges due to shifts in core power distribution. To address these complexities, a multi-physics approach is employed using the multi-physics package RELAP5/3DKIN and implementing two strategies. The first strategy uses a mechanical shim, adjusting the reactor power exclusively through CEAs. The second strategy combines CEA movement with adjustments in soluble boron concentration. Both strategies are evaluated against axial offset and 3D power peaking safety limits to ensure compliance with operational safety requirements. Full article

The analysis of liquid-fueled molten-salt reactors (LFMSRs) during steady state, operational transients and accident scenarios requires addressing unique reactor multi-physics challenges with coupling between thermal hydraulics, neutronics, inventory control and species distribution phenomena. This work utilizes the General Nuclear Field Operation and Manipulation (GeN-Foam) code to perform coupled thermal-hydraulics and neutronics calculations of an LFMSR design. A framework is proposed as part of this study to perform modeling, design optimization, and uncertainty quantification. The framework aims to establish a protocol for the studies and analyses of LFMSR which can later be expanded to other advanced reactor concepts too. The Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) statistical analysis tool was successfully coupled with GeN-Foam to perform uncertainty quantification studies. The uncertainties were propagated through the input design parameters, and the output uncertainties were characterized using statistical analysis and Spearman rank correlation coefficients. Three analyses are performed (namely, scalar, functional, and three-dimensional analyses) to understand the impact of input uncertainty propagation on temperature and velocity predictions. Preliminary three-dimensional reactor analysis showed that the thermal expansion coefficient, heat transfer coefficient, and specific heat of the fuel salt are the crucial input parameters that influence the temperature and velocity predictions inside the LFMSR system. Full article

Nuclear graphite plays a crucial role in thermal-spectrum thorium molten salt reactors (TMSRs) as both the neutron moderator and the construct for the coolant flowing channel. When subjected to irradiation and elevated temperatures, graphite components experience considerable deformation due to a combination of dimensional changes, thermal expansion, irradiation creep, elastic deformation, and changes in thermomechanical characteristics. The lifespan of the graphite component is a limiting factor in TMSR designs as it strongly correlates with the dimensional changes of the graphite. To evaluate the thermal and mechanical reactions of graphite component under TMSR core conditions, it is necessary to couple models of thermal-hydraulics, neutronics, and thermal-mechanics. This paper presents an enhanced methodology for analyzing the deformation of graphite components using the finite element method. Then, this method was applied to analyze a 10-year deformation history of a hexagonal prism assembly (HPA) and it was compared with the traditional hexagonal round channel assembly (RCA). The results demonstrate that the stress–strain field of both types of graphite components undergo significant variations with the increasing neutron fluence from irradiation. HPA graphite exhibits a slower deformation as compared to RCA graphite when subjected to identical operating conditions. In this case, HPA graphite has a lifespan of approximately 10 years, while RCA graphite lasts only 8.8 years. Full article

Jacobian-free Newton Krylov (JFNK) is an attractive method to solve nonlinear equations in the nuclear engineering community, and has been successfully applied to steady-state neutron diffusion k-eigenvalue problems and multi-physics coupling problems. Preconditioning technique plays an important role in the JFNK algorithm, significantly affecting its computational efficiency. The key point is how to automatically construct a high-quality preconditioning matrix that can improve the convergence rate and perform the preconditioning matrix factorization efficiently and robustly. A reordering-based ILU(k) preconditioner is proposed to achieve the above objectives. In detail, the finite difference technique combined with the coloring algorithm is utilized to automatically construct a preconditioning matrix with low computational cost. Furthermore, the reordering algorithm is employed for the ILU(k) to reduce the additional non-zero elements and pursue robust computational performance. A 2D LRA neutron steady-state benchmark problem is used to evaluate the performance of the proposed preconditioning technique, and a steady-state neutron diffusion k-eigenvalue problem with thermal-hydraulic feedback is also utilized as a supplement. The results show that coloring algorithms can automatically and efficiently construct the preconditioning matrix. The computational efficiency of the FDP with coloring could be about 60 times higher than that of the preconditioner without the coloring algorithm. The reordering-based ILU(k) preconditioner shows excellent robustness, avoiding the effect of the fill-in level k choice in incomplete LU factorization. Moreover, its performances under different fill-in levels are comparable to the optimal computational cost with natural ordering. Full article

Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model to properly account for the radial leakage effects of the reactor. The leakage cross section and other consistent neutronics parameters are generated with the Monte Carlo code Serpent using high-fidelity three-dimensional (3D) models. The accuracy of the 1D consistent model is verified by the reference solution from the Monte Carlo model on the Molten Salt Reactor Experiment (MSRE) configuration. The 1D consistent model successfully reproduced the integrated flux from the 3D model and the reactor multiplication factor k_eff with the error in the range of 95 to 397 pcm (per cent mille), depending on discretized energy group structures. The developed model is also extended to estimate the reactivity loss due to fuel circulation in MSRE. The estimate of reactivity loss in dynamics analysis is in great agreement with the experimental data. This model functions as the first step in the development of a 1D fully neutronics/thermal-hydraulics coupled model for short- and long-term MSRE transient analysis. Full article

The Xi’an Pulsed Reactor (XAPR) is characterized by its small core size and integrated fuel moderator structure, which results in a non-uniform core power and temperature distribution. Consequently, a complex coupling relationship exists between its core neutronics and thermal hydraulics, necessitating the assurance for the operational safety of the XAPR. To optimize the experimental scheme in the reactor, a refined three-dimensional steady-state nuclear-thermal coupling analysis is imperative. This study focuses on investigating the coupling calculation of a three-dimensional steady-state neutronics and thermal-hydraulics model for the XAPR by utilizing an open-source multi-physical coupling framework known as Cardinal. The neutron transport equation is effectively solved using OpenMC, while a three-dimensional heat conduction model is employed to compute the heat conduction of the fuel elements. Furthermore, a parallel multi-channel model is utilized to determine the fluid heat transfer. The research is centered on the XAPR, whereby Monte Carlo and thermal-hydraulics coupling calculations of the core under steady-state full-power conditions are conducted, specifically at an operational capacity of 2 MW. The results demonstrate a strong agreement between the simulation and experimental outcomes. The maximum temperature recorded for the thermometric fuel element in the XAPR is 795.1 K, with a deviation of approximately −5.7% from the measured value. Moreover, the outlet fluid temperature of the thermal channel is observed to be 360 K, exhibiting a deviation of around −2.7% from the measured value. Full article

Molten salt reactors (MSRs) are one type of GEN-IV advanced reactors that adopt melt mixtures of heavy metal elements and molten salt as both fuel and coolant. The liquid fuel allows MSRs to perform online refueling, reprocessing, and helium bubbling. The fuel utilization, safety, and economics can be enhanced, while some new physical mechanisms and phenomena emerge simultaneously, which would significantly complicate the numerical simulation of MSRs. The dual roles of molten fuel salt in the core lead to a tighter coupling of physical mechanisms since the released fission energy will be absorbed immediately by the molten salt itself and then transferred to the primary heat exchanger. The modeling of multi-physics coupling is regarded as one important aspect of MSR study, attracting growing attention worldwide. Up to now, great efforts have been made in the development of MSR multi-physics coupling models over the past 60 years, especially after 2000, when MSR was selected for one of the GEN-IV advanced reactors. In this paper, the development status of the MSR multi-physics coupling model is extensively reviewed in the light of coupling models of N-TH (neutronics and thermal hydraulics), N-TH-BN (neutronics, thermal hydraulics, and burnup) and N-TH-BN-G (neutronics, thermal hydraulics, burnup, and graphite deformation). The problems, challenges, and development trends are outlined to provide a basis for the future development of MSR multi-physics coupling models. Full article

Nuclear thermal propulsion is an enabling technology for future space missions, such as crew-operated Mars missions. Nuclear thermal propulsion technology provides a performance benefit over chemical propulsion systems by operating with light propellants (e.g., hydrogen) at elevated engine chamber conditions. Therefore, nuclear thermal propulsion reactor cores exhibit high propellant velocities and elevated propellant and fuel temperatures, subsequently leading to relatively high thermal stresses and geometrical deformation. This paper details the numerical approach to solve the thermo-elastic equations, which was implemented into the recently developed ntpThermo code. In addition, this paper demonstrates the extension of the Basilisk multiphysics framework to perform full-core coupled neutronic, thermal-hydraulic, and thermo-mechanical analysis of nuclear thermal propulsion reactors. The analyses demonstrate and quantify thermo-mechanical feedback, which for the investigated cases, acted to reduce maximum fuel temperatures and pressure drop across the fuel element channels. Thermo-mechanical feedback had a significant impact on the mass flow distribution within the reactor core and, thus, a substantial impact on solid-material temperatures and stresses, but only a minor impact on reactivity and local power distributions. Sensitivity studies revealed that the friction factor correlation applied to perform the analysis has a significant impact on the pressure drop across the fuel element channels. The most important observation of this research is the importance of incorporating the thermo-mechanical feedback within an integrated multiphysics solution sequence to enable the consistent design of future nuclear thermal propulsion systems. Full article

Best Estimate Plus Uncertainty (BEPU) approaches for nuclear reactor applications have been extensively developed in recent years. The challenge for BEPU approaches is to achieve multi-physics modeling with an acceptable computational cost while preserving a reasonable fidelity of the physics modeled. In this work, we present the core multi-physics computational framework developed for the efficient computation of uncertainties in Light Water Reactor (LWR) simulations. The subchannel thermal-hydraulic code CTF and the nodal expansion neutronic code PARCS are coupled for the multi-physics modeling (CTF-PARCS). The computational framework is discussed in detail from the Polaris lattice calculations up to the CTF-PARCS coupling approaches. Sampler is used to perturb the multi-group microscopic cross-sections, fission yields and manufacturing parameters, while Dakota is used to sample the CTF input parameters and the boundary conditions. Python scripts were developed to automatize and modularize both pre- and post-processing. The current state of the framework allows the consistent perturbation of inputs across neutronics and thermal-hydraulics modeling. Improvements to the standard thermal-hydraulics modeling for such coupling approaches have been implemented in CTF to allow the usage of 3D burnup distribution, calculation of the radial power and the burnup profile, and the usage of Santamarina effective Doppler temperature. The uncertainty quantification approach allows the treatment of both scalar and functional quantities and can estimate correlation between the multi-physics outputs of interest and up to the originally perturbed microscopic cross-sections and yields. The computational framework is applied to three exercises of the LWR Uncertainty Analysis in Modeling Phase III benchmark. The exercises cover steady-state, depletion and transient calculations. The results show that the maximum fuel centerline temperature across all exercises is $2474\mathrm{K}$ with $1.7\%$ uncertainty and that the most correlated inputs are the ²³⁸U inelastic and elastic cross-sections above 1 MeV. Full article