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        <title>Frontiers in Nuclear Engineering | Nuclear Reactor Design section | New and Recent Articles</title>
        <link>https://www.frontiersin.org/journals/nuclear-engineering/sections/nuclear-reactor-design</link>
        <description>RSS Feed for Nuclear Reactor Design section in the Frontiers in Nuclear Engineering journal | New and Recent Articles</description>
        <language>en-us</language>
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        <pubDate>2026-05-13T22:17:40.975+00:00</pubDate>
        <ttl>60</ttl>
        <item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2026.1776967</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2026.1776967</link>
        <title><![CDATA[Enhancing fast neutron irradiation in thermal neutron spectrum reactors through python-based multi-objective optimization]]></title>
        <pubdate>2026-05-08T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Mustafa K. Jaradat</author><author>Jason V. Brookman</author><author>Andrew Bascom</author><author>Tommy Holschuh</author><author>Nicolas E. Woolstenhulme</author>
        <description><![CDATA[This work reports on a pilot study for optimizing the design of a fast neutron irradiation experiment in a thermal neutron spectrum, specifically the Advanced Test Reactor (ATR). A fast and robust multi-objective optimization workflow that leverages Python-based open-source tools was developed and applied to the ATR to optimize experiment design and boost fast energy neutrons at a desired irradiation location. Three design options were explored to minimize thermal and epithermal neutron flux, deposited heat, and total estimated cost while maximizing the absolute fast neutron flux. This was achieved by considering several irradiation positions in the ATR with different combinations and thicknesses of filter and booster materials. The developed workflow utilizes high-fidelity Monte Carlo calculations to train a surrogate model of each objective function being optimized, thereby reducing computational efforts while searching for the optimized set of solutions. The results show that absolute fast neutron flux increased approximately 30% to 55% in regions with a harder spectrum, while the absolute fast neutron flux increased significantly by 7 to 10 times in regions with a softer spectrum outside the core but still lower than the regions with harder spectrum. Also, The predictions of the surrogate models were verified against the high-fidelity Monte Carlo calculations, and these tests showed that the surrogate models made accurate predictions.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2026.1771702</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2026.1771702</link>
        <title><![CDATA[Research on a strongly generalizable fault diagnosis method based on adversarial transfer learning]]></title>
        <pubdate>2026-04-13T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Biwei Zhu</author><author>Zhiguang Deng</author><author>Xuemei Wang</author><author>Sijie Xu</author><author>Chenlong Dong</author>
        <description><![CDATA[IntroductionShallow machine learning algorithms exhibit low efficiency in fault diagnosis under the conditions of small-sample and unlabeled data. To address this critical problem, this paper focuses on developing an effective fault diagnosis method suitable for cross-reactor-type scenarios, which is of great significance for improving the safety and operational level of nuclear power plants.MethodsA cross-reactor-type fault diagnosis method based on adversarial transfer learning is proposed. By integrating deep learning and transfer learning techniques, a hybrid domain-adversarial learning model is constructed. The overall loss function of the model is designed to effectively extract transferable features between related reactor types, and corresponding validation experiments are carried out to verify the model's feasibility and effectiveness.ResultsThe experimental validation shows that the proposed hybrid domain-adversarial learning model can effectively extract transferable features across different reactor types, which solves the problem of low efficiency of shallow machine learning algorithms in fault diagnosis under small-sample and unlabeled data conditions. The model achieves reliable fault diagnosis performance in cross-reactor-type scenarios.DiscussionWhen applied to cross-reactor-type nuclear power plant fault diagnosis, the research findings can significantly enhance the safety of nuclear power plants, improve their economic performance and operational efficiency. Furthermore, this research effectively promotes the intelligence level and autonomous decision-making capabilities of nuclear power plants, providing a valuable technical reference for the intelligent development of the nuclear power industry.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2026.1771859</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2026.1771859</link>
        <title><![CDATA[Validation of SIMULATE5-K and CASMO5 with the SPERT-III E-Core]]></title>
        <pubdate>2026-03-20T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>William Dawn</author><author>Gerardo Grandi</author><author>Tamer Bahadir</author>
        <description><![CDATA[SIMULATE5-K (S5K) is the newest state-of-the-art, best-estimate, transient reactor analysis software developed by Studsvik Scandpower, Inc. (SSP). When used in conjunction with CASMO5 to generate multi-group neutronic data for steady-state and transient calculations, S5K can be used to accurately model reactor transient conditions in Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). Recently, the methods in S5K and CASMO5 were validated by simulating reactor transients from the SPERT-III E-core experiments. These experiments were performed on a small Light Water Reactor (LWR) core and were designed to resemble Reactivity Insertion Accidents (RIAs) in PWRs. Overall, the results calculated with S5K agree well with the experimentally measured data and any differences are much smaller than the reported uncertainties in the measurements, especially the uncertainty in the initial condition. Additionally, S5K supports general multi-group time-dependent neutron diffusion calculations and the SPERT-III E-core experiments were used to show that there were no obvious trends or discrepancies when the two-, four-, and eight-group calculations were compared.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2026.1714531</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2026.1714531</link>
        <title><![CDATA[Evaluation of the Lawson criterion for aneutronic proton-boron-11 fusion: effects of ion temperature and bremsstrahlung losses]]></title>
        <pubdate>2026-02-24T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Irfan Maulana Ahmad</author><author>Abd. Djamil Husin</author><author>Duong Thanh Tai</author><author>Nissren Tamam</author><author>Abdelmoneim Sulieman</author><author>Sitti Yani</author>
        <description><![CDATA[Nuclear fusion, the process of combining light nuclei to form heavier nuclei, offers a promising pathway to sustainable clean energy with minimal radioactive waste. The Lawson criterion, expressed as the product of plasma density, confinement time, and temperature, establishes the conditions required for ignition and net energy gain. This study investigates the Lawson criterion for proton-boron-11 (p-11B) fusion across ion temperatures of 75–500 keV, incorporating fusion reactivity data from Tentori-Belloni and Nevins-Swain, as well as energy losses from bremsstrahlung radiation under different electron-to-ion temperature ratios (TeTi= 1, 0.5, 0.25). The Tentori-Belloni dataset yields higher fusion reactivity than Nevins-Swain, resulting in more favorable Lawson values. Net energy production is achieved only when Te<Ti, with optimal operating windows identified at 190–330 keV for Te= 0.5Ti and 125–500 keV for Te= 0.25Ti. At Ti< 230 keV, the Lawson criterion decreases due to plasma instabilities and confinement limitations; in this work, radiative losses are evaluated using Zeff=2.4 derived from the p-11B fuel mixture (npnB=90:10) only, while external impurity contributions are not explicitly modeled. For Ti> 230 keV, the Lawson criterion increases, reaching characteristic minima around 330 keV and 500 keV. These thresholds represent the minimum conditions required to achieve ignition and sustain a self-sufficient fusion reaction. The minimum Lawson values obtained were 1.3 × 1022 m−3s (no radiation), 1.2 × 1023 m−3s (Te= 0.5Ti), and 1.5 × 1022 m−3s (Te= 0.25Ti). These findings highlight the critical role of accurate cross-section data and electron-ion temperature control in advancing aneutronic p-11B fusion toward practical, self-sustained clean energy systems.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1720142</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1720142</link>
        <title><![CDATA[High-fidelity multi-physics guidelines for model validation and uncertainty quantification]]></title>
        <pubdate>2026-02-04T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Gregory K. Delipei</author><author>Quentin Faure</author><author>Maria Avramova</author><author>Kostadin Ivanov</author>
        <description><![CDATA[The verification, validation, and uncertainty quantification (VVUQ) of high-fidelity, high-resolution multi-physics modeling and simulation in nuclear engineering applications are essential for assessing the predictive credibility of developed models. Appropriate practices and methods are required to address ongoing challenges. Some key examples include the large dimensionality of the input and output spaces, modeling complexity, high computational cost, scarcity of relevant experimental data, and the lack of guidelines and protocols for the development of multi-physics benchmarks. This study provides several guidelines and recommendations. Dimensionality reduction and screening approaches can be used to address the high-dimensional input and output spaces. A multi-level validation hierarchy where the coupling level is increased progressively is suggested to manage modeling complexity. A validation scoring method is proposed to compare the different coupling levels and to identify gaps in the modeling. Surrogate models can be used to address the computational cost, though they require the estimation of an additional model uncertainty. For consistent uncertainty propagation, sample-processing diagrams are introduced that can help avoid sampling errors between the multiple inputs. For the validation of multivariate outputs such as time series, local, regional, and global univariate metrics can be used together with more complicated multivariate methods based on U-pooling. Some of the proposed recommendations are demonstrated on the multi-physics modeling of the first cold ramp test from the OECD/Nuclear Energy Agency (NEA) Multi-physics Pellet Cladding Mechanical Interaction Validation (MPCMIV) benchmark. The multi-level modeling hierarchy ranges from single-physics fuel performance models to coupled multi-physics models. The MOOSE-based tools Griffin, Bison, and THM are employed alongside the fuel performance code OFFBEAT. The measurements considered in here include the cladding’s axial elongation and coolant temperature at three different locations during the cold ramp test. Validation metrics are computed at local, regional, and global scales. Validation scores are computed for each model and physics domain. The results highlight the need for at least a coupling between the RP and FP to accurately predict the cladding axial elongation, whereas the coolant temperatures are less sensitive to the coupling level due to their small variations during the cold ramp test.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1694684</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1694684</link>
        <title><![CDATA[Accelerating the engineering design of breeder blankets with parametric optimisation and sequential learning]]></title>
        <pubdate>2026-02-02T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Luke Humphrey</author><author>Helen Brooks</author><author>Siddharth Mungale</author><author>Andrew Davis</author><author>David Foster</author>
        <description><![CDATA[The competing requirements of fusion breeder blankets and the high dimensionality of their design space necessitate a systematic treatment to map the variations in performance against given objective metrics and to understand the operational envelope. In this endeavour, a digital engineering pipeline for design evaluation and optimisation has been developed. The tools involved are Hypnos for parametric breeder blanket geometry instantiation, OpenMC for neutronics analysis, MOOSE for thermal hydraulics analysis, and SLEDO for design space sampling, sensitivity analysis, and optimisation. An optimisation of the baseline design for a solid ceramic breeder mock-up that is relevant to the Lithium Breeding Tritium Innovation (LIBRTI) program is performed. Two optimisation studies are performed, the first involving only neutronics, while the second includes the impact of thermal hydraulics. The figures of merit are taken to be the tritium breeding ratio (TBR) and the pressure drop of the outer coolant (combined in a weighted sum for the second analysis). In the first study, for the same acquisition function (taken to be expected improvement), two different values are selected for the hyperparameter that controls the trade-off between exploration and exploitation. In the second study, with the inclusion of thermal hydraulics, a larger parameter space was explored to assess the performance of the method in a higher dimensionality setting. In both cases, the selected figures of merit were improved over the baseline design. Finally, we discuss extensions of the procedure to include a more thorough multi-physics analysis and a more sophisticated treatment of multiple objectives.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2026.1762086</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2026.1762086</link>
        <title><![CDATA[Pebble dynamics and thermal-fluid analysis of high-temperature gas-cooled pebble bed reactors using DEM and CFD simulations]]></title>
        <pubdate>2026-01-30T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Kashminder S. Mehta</author><author>Braden Goddard</author><author>Zeyun Wu</author>
        <description><![CDATA[This study presents a multiphysics computational simulation framework for analyzing pebble dynamics and thermal-fluid behavior in High-Temperature Gas-Cooled Pebble Bed Reactors (HTG-PBR). The pebble circulation and intermixing effects are predicted using Discrete Element Method (DEM) implemented in LIGGGHTS, while the thermal-fluid behavior is simulated with computational fluid dynamics (CFD) in OpenFOAM. The CFD model employs a porous-media formulation with a local thermal non-equilibrium model to capture the energy exchange between the helium coolant and pebbles. Integrating the DEM-based mixing effects into the porous CFD model enables a more physically representative and scalable approach for full-core reactor analysis. Both DEM and CFD solvers are validated using established pebble-bed benchmark problems to confirm the viability of the developed computational models. A HTG-PBR-like conical model reactor is employed as a test problem to evaluate the developed method. The simulation results confirm the predictive capability of the developed models for HTG-PBR performance analysis and provide insight for future multiphysics coupling strategies for reactor design optimization.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1751259</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1751259</link>
        <title><![CDATA[Correction: Editorial: Multiphysics methods and analysis applied to nuclear reactor systems]]></title>
        <pubdate>2025-12-17T00:00:00Z</pubdate>
        <category>Correction</category>
        <author>Mark D. DeHart</author><author>Emily Shemon</author><author>Deokjung Lee</author>
        <description></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1677436</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1677436</link>
        <title><![CDATA[Multi-step verification of the Copenhagen atomics molten salt experiment radioactive inventory at the Paul Scherrer Institute using OpenMC, Serpent, and EQL0D]]></title>
        <pubdate>2025-12-08T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Lorenzo Chierici</author><author>Alessio Rossi</author>
        <description><![CDATA[The molten salt experiment (MSE) is a 30-day 1 MW (thermal) molten salt criticality experiment designed by Copenhagen Atomics that will be conducted at the Paul Scherrer Institute (PSI) Swiss research center. A detailed analysis of the radioactive inventory available during the operation and after shutdown has been carried out for future estimates of potential source term release in case of undesired accidents using the Monte Carlo transport and depletion code OpenMC. To benchmark the radioactive inventory obtained through the OpenMC burn-up calculation (version 0.15.2), code-to-code comparison with the Serpent Monte Carlo and depletion code (v2.2.1), as well as with Serpent coupled to the EQL0D depletion module developed at PSI, were performed. All of the utilized codes support neutron transport on CAD-based geometries and allow continuous material reprocessing during depletion. This feature is essential for modeling the MSE concept, where gaseous fission products are continuously extracted from the circulating fuel salt. For the present verification study, three models of increasing geometric and physical detail were developed: (1) a nested hollow sphere model in constructive solid geometry (CSG), with the scope of aligning basic settings beyond geometry; (2) a geometrically simplified CAD model of the Copenhagen Atomics Onion Core, mainly utilized for testing the integration of CAD models within the above depletion codes; (3) the same simplified CAD model but with the implementation of reactor fluid flows from the inner active core to the outer inactive region, mainly to account for the effect of out-of-core decay. The results were compared to benchmark the original OpenMC model. Particular attention was given to the end-of-cycle (EOC) isotopic inventory, with detailed comparisons for key radionuclides and total radiotoxicity.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1717262</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1717262</link>
        <title><![CDATA[Editorial: Multiphysics methods and analysis applied to nuclear reactor systems]]></title>
        <pubdate>2025-10-24T00:00:00Z</pubdate>
        <category>Editorial</category>
        <author>Mark D. DeHart</author><author>Emily Shemon</author><author>Deokjung Lee</author>
        <description></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1597165</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1597165</link>
        <title><![CDATA[Hybrid solvers for reactor modelling: matrix-based and matrix-free approaches on voxel-dominated meshes]]></title>
        <pubdate>2025-10-15T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Liang Yang</author><author>Jianhui Yang</author><author>Jakov Popov</author><author>Andrew G. Buchan</author>
        <description><![CDATA[Simulating neutronics and thermal hydraulics within nuclear reactor cores is computationally intensive, not only because of the complexity of the governing equations but also because of the intricate geometries involved. Solving the Boltzmann transport and Navier-Stokes equations for a full core representation typically relies on unstructured meshes, which, while highly flexible, can substantially increase computational costs regarding memory and solving time. Cartesian meshes with Finite Elements (FE) offer a faster alternative, potentially improving computational speed by an order of magnitude due to direct memory addressing. However, they necessitate finer grids to accurately capture the boundary details of non-Cartesian surfaces, which can offset these gains by increasing solver times. To address this challenge, a new meshing algorithm is proposed in conjunction with hybrid, matrix-based and matrix free, solver technologies. It employs a geometry-conforming boundary method using voxel-dominated Cartesian meshes. This method enables accurate boundary representation at arbitrary resolutions, which can be adjusted to resolve the physics to the desired level of accuracy rather than strictly to capture geometric detail. This is combined with a hybrid solver for fluid flows to different regions of a problem in order to increase efficiency when resolving the boundary. This article demonstrates the method’s application to Computational Fluids Dynamics (CFD) and neutronics problems relevant to reactor physics, showcasing its accuracy, convergence, numerical stability, and suitability for handling complex geometries.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1628866</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1628866</link>
        <title><![CDATA[Uncertainty quantification and sensitivity analysis of a nuclear thermal propulsion reactor startup sequence]]></title>
        <pubdate>2025-10-08T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Jackson R. Harter</author><author>Mark D. DeHart</author>
        <description><![CDATA[The research presented in this article describes progress in applying stochastic methods, uncertainty quantification, parametric studies, and variance-based sensitivity analysis (also known as Sobol sensitivity analysis) to a full-core model of a nuclear thermal propulsion (NTP) system simulated via the radiation transport code Griffin to simulate neutronics. Our goal is to develop a reduced-order (surrogate) model that can be rapidly sampled with perturbations to multiple input parameters. In this NTP system, reactivity and power feedback affect the rotation of control drums (CDs), which is itself controlled by a hybrid proportional-integral-derivative (PID) controller actuated by the power demand and reactivity feedback from the numerical model. This model uses reactor kinetic feedback (mean generation time [Λ] and effective delayed neutron fraction [βeff] from a transient Griffin simulation executed via Griffin’s improved quasi-static solver to provide the kinetic parameters) as inputs to functions that control the CD rotation angle. By investigating numerous stochastic approaches, we developed a dual-purpose surrogate model of the NTP system, using polynomial regression in the Multiphysics Object-Oriented Simulation Environment (MOOSE) Stochastic Tools Module (STM). The trained model can be rapidly sampled while simultaneously perturbing various input parameters, such as coefficients on the PID control or temperature (directly affecting the neutron cross section). The surrogate model delivers accurate (within 5%) results at speeds orders of magnitude faster (minutes, not days of computational time) than the base model. Once the surrogate model has been trained, distributions of the uncertain parameters can be changed at will to investigate the effects of perturbing multiple inputs as well as the effects of these inputs on the model output. For example, coefficients used in the PID control system may vary due to some type of physical interference, or uncertainty may exist in the temperature of the neutron cross sections in various regions of the reactor. A distribution can be placed on these parameters, and operational boundaries can be determined. The goal of this work is to support development of an advanced control system for operating CDs in a functioning NTP system. This work is a scoping study of the MOOSE STM.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1691453</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1691453</link>
        <title><![CDATA[Editorial: Nuclear reactor safety and accident mitigation management]]></title>
        <pubdate>2025-09-17T00:00:00Z</pubdate>
        <category>Editorial</category>
        <author>Amit Kumar</author><author>Sanjeev Gupta</author><author>Yuh-Ming Ferng</author>
        <description></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1620419</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1620419</link>
        <title><![CDATA[An overview of research groups and their activities in the field of inertial electrostatic confinement fusion]]></title>
        <pubdate>2025-08-25T00:00:00Z</pubdate>
        <category>Review</category>
        <author>Jan-Philipp Wulfkühler</author><author>Mahmoud Bakr</author><author>Martin Tajmar</author>
        <description><![CDATA[Inertial electrostatic confinement fusion has developed into a widespread academic field since its inception in the 1950s and 1960s. This paper provides an overview of the different research groups (universities and research institutes) and companies involved in the field of IECF and their scientific publications. A list of over 970 publications from 56 universities, 20 research institutes, and 25 companies was collected and analyzed. Also, an overview of the most common type of IECF devices, often referred to as “gridded” IECF device or “fusor” was created, including more than 30 devices. This paper serves as both a reference guide to the literature and the IECF devices.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1617048</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1617048</link>
        <title><![CDATA[Comparison of spatial dynamics and point kinetics approaches in multiphysics modeling of the molten salt reactor experiment]]></title>
        <pubdate>2025-08-11T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Philip Pfahl</author><author>Mustafa K. Jaradat</author><author>Mauricio E. Tano</author><author>Ramiro O. Freile</author><author>Samuel A. Walker</author><author>Javier Ortensi</author>
        <description><![CDATA[In this work, we present validation test results of fully coupled neutronics and thermal-hydraulics models of the Molten Salt Reactor Experiment (MSRE) against experimental data of the zero power pump transients and the natural circulation tests at low power. To capture the strong coupling between neutronics and thermal-hydraulics due to fuel circulation, and to account for the delayed neutron precursor (DNP) distribution, the porous media thermal-hydraulics solver Pronghorn was fully coupled to the spatial neutron dynamics code Griffin, which solves the neutron diffusion equation, and to the 0-D point kinetics solver Squirrel, using a 2-D homogenized representation of the MSRE. The validation test results show very good agreement with experimental data for both point kinetics and spatial dynamics simulations, capturing the strong feedback effect and DNP losses in the MSRE. The 0-D code Squirrel accurately predicted the time-dependent behavior in the MSRE given the steady-state spatial dynamics solution of Griffin.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1594698</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1594698</link>
        <title><![CDATA[CTF development, verification, and validation for VVER core thermal-hydraulics and multi-physics modeling and simulation]]></title>
        <pubdate>2025-08-01T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Yesim Kutlu</author><author>Ivan Spasov</author><author>Svetlomir Mitkov</author><author>Pascal Rouxelin</author><author>Agustin Abarca</author><author>Nikola Kolev</author><author>Maria Avramova</author><author>Kostadin Ivanov</author>
        <description><![CDATA[The advanced thermal-hydraulics sub-channel tool CTF has been in the process of continuous development and improvement by Oak Ridge National Laboratory (ORNL) and North Carolina State University (NCSU). In recent years, there has been considerable progress in code development, including new functionalities, application-specific correlations, various multi-physics applications, built-in pre- and post-processors, improved solvers, parallelization, and extensive testing. VVER applications are part of these activities. NCSU has been cooperating with the Institute for Nuclear Research and Energy (INRNE) on CTF development, verification, and validation for VVER core modeling and simulation. This article presents an overview of these CTF studies for VVER applications. Several test cases are considered, which include pure thermal-hydraulic problems as well as multi-physics simulations at the nodal and pin level. On the single physics side, thermal-hydraulic CTF solutions have been compared against measured data for rod bundle, fuel assembly, and full core, as well as code-to-code vs. FLICA4 solutions. CTF was tested in the simulation of the TVSA-5T VVER mini-assembly experiments and in the full-core steady-state calculation for the ongoing OECD/NEA Rostov-2 benchmark. For the TVSA-5T calculations, CTF was coupled with the uncertainty analysis tool Dakota and utilized to propagate uncertainties of input and boundary conditions to output quantities of interest for thermal-hydraulic parameter investigations. The CTF results and measured data obtained from this experimental setup were compared for validation. To produce reliable pin-resolved reference solutions for multi-physics model testing the high-fidelity continuous energy Monte Carlo-based neutron transport codes MCNP6.2 and Serpent 2.2.0 were separately coupled with the CTF sub-channel code. Coupled models of a VVER-1000 fuel assembly were tested in comparisons between MCNP/CTF and Serpent/CTF results. Coarse-mesh multi-physics solutions for a full core have been obtained with the coupled COBAYA/CTF, COBAYA/FLICA4, and PARCS/CTF codes. These solutions have been compared against steady-state plant data and code-to-code for transients. High-fidelity pin-resolved solutions with SERPENT/CTF serve as reference solutions in a steady state. The outcomes from the various studies of single-physics and multi-physics cases used for CTF verification and validation met the initial expectations both qualitatively and quantitatively. The results of the numerical verification and experimental validation are in good agreement with the corresponding reference data.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1611173</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1611173</link>
        <title><![CDATA[Data transfers for nuclear reactor multiphysics studies using the MOOSE framework]]></title>
        <pubdate>2025-07-25T00:00:00Z</pubdate>
        <category>Methods</category>
        <author>Guillaume L. Giudicelli</author><author>Fande Kong</author><author>Roy Stogner</author><author>Logan Harbour</author><author>Derek Gaston</author><author>Alexander Lindsay</author><author>Zachary Prince</author><author>Lise Charlot</author><author>Stefano Terlizzi</author><author>Mahmoud Eltawila</author><author>April Novak</author>
        <description><![CDATA[High fidelity simulations of nuclear systems generally require a multi-dimensional representation of the system. Advanced nuclear reactor cores are governed by multiple physical phenomena which should be all be resolved, and the coupling of these physics would also need to be resolved spatially in a high-fidelity approach, while lower fidelity may leverage integrated quantities for the coupling instead. Performing a spatially resolved multiphysics simulation can be done on a single mesh with a single coupled numerical system, but this requires catering to each equations’ time and spatial discretization needs. Instead, each physics, usually neutronics, thermal hydraulics and fuel performance, are solved individually with the discretization they require, and the equations are coupled by transferring fields between each solver. In our experience coupling applications within the MOOSE framework, mostly for advanced nuclear reactor analysis, there are several challenges to this approach, from non-conservation problems with dissimilar meshes, to losses in order of spatial accuracy. This paper presents the field transfer capabilities implemented in MOOSE, and numerous technical details such as mapping heuristics, conservation techniques and parallel algorithms. Examples are drawn from nuclear systems analysis cases to illustrate the techniques.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1595628</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1595628</link>
        <title><![CDATA[Coupling of the Monte Carlo iMC and OpenFoam codes for multiphysics calculations of molten salt reactors]]></title>
        <pubdate>2025-07-11T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Inyup Kim</author><author>Taesuk Oh</author><author>Yonghee Kim</author>
        <description><![CDATA[This paper presents the development of a multiphysics coupled framework of Monte Carlo neutronics iMC and OpenFOAM Computational Fluid Dynamics codes for molten salt reactor (MSR) analysis. The overall coupling scheme is handled, including the framework structure and iteration scheme. Also, related techniques to enhance the accuracy and efficiency of the coupling are introduced, such as delayed neutron precursor tracking. The framework is applied to a simple molten salt reactor model and achieves a converged solution. In addition, sensitivity studies on the neutronics mesh are performed. The research demonstrates the capability of the iMC-OpenFOAM coupled framework to achieve a converged solution and provides significant insights into the analysis of the MSRs.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1537136</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1537136</link>
        <title><![CDATA[Improved high-fidelity multiphysics modeling of pulsed operation of the annular core research reactor]]></title>
        <pubdate>2025-06-04T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Emory Colvin</author><author>Todd S. Palmer</author>
        <description><![CDATA[Sandia National Laboratories’ Annular Core Research Reactor (ACRR) is a unique research reactor, using UO2-BeO fuel and operating primarily in pulsed mode. To better understand the physical characteristics of the fuel, the distribution of heat generation must be understood. Previous work developed a Serpent two model of the ACRR and Python coupling script to provide multiphysics feedback. Simulations of $1.50 and $2.00 pulses were compared to experimental results. This paper expands this work to $2.50 and $3.00 pulses. It further explores potential improvements to the model: dividing the fuel into two radial regions for feedback purposes, allowing additional iterations of the multiphysics coupling and checking for convergence, and the development of alternate specific heat capacity values. The use of two radial fuel regions improved agreement with experimental results for the simulations using the original function for specific heat capacity as a function of temperature but did not consistently improve results with the constant value for specific heat capacity. Allowing additional multiphysics iterations until the power distribution field converges also produced little change for reactor power prediction, though it improved maximum fuel temperature prediction slightly. The new values for specific heat capacity provided the most significant improvements to the models. A third-order polynomial developed from experimental data results in a significant improvement in fuel temperature prediction over the constant value with only a small loss of performance in reactor power prediction.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1589780</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1589780</link>
        <title><![CDATA[Coupled neutronic-thermal-mechanical simulation of the KRUSTY heat pipe microreactor]]></title>
        <pubdate>2025-04-29T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>William Reed Kendrick</author><author>Benoit Forget</author>
        <description><![CDATA[Multiphysics analysis has become a common technique for nuclear reactor design validation, with neutronic-thermal analysis being the typical choice for understanding reactor dynamics. The concept of adding mechanical simulation such as thermal expansion to this coupling is still relatively new, however, and presents many computational challenges. While large reactors see relatively little neutronic impact from thermal expansion and may not warrant the challenge of undertaking this level of coupling, recent studies of microreactor geometries show that smaller reactors see larger impacts from thermal expansion. This work performs coupled neutronic-thermal-mechanical simulation of the Kilowatt Reactor Using Stirling TechnologY (KRUSTY) using OpenMC and Multiphysics Object-Oriented Simulation Environment in order to analyze the neutronic and thermal impact of including thermal expansion at steady state. The results show that while thermal expansion has a significant effect on global neutronic tallies, it has relatively minor impact on spatial heating rates or temperatures in the system. This remains true even when simulating a multiple heat pipe failure scenario to introduce thermal asymmetry.]]></description>
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