Pub Date : 2026-01-12DOI: 10.1016/j.nucengdes.2025.114727
Ji-Woong Han, Sun Rock Choi, In Sub Jun, Seungjoon Baik, Huee-Youl Ye, Jewhan Lee
This paper investigates the diversity of trip parameters in the reactor protection system (RPS) of SALUS, a Small, Advanced, Long-cycled, and Ultimate safe Sodium-cooled fast reactor, for various anticipated operational occurrences (AOOs). Three representative AOO events are selected: transient overpower (TOP), loss of heat sink (LOHS), and loss of core flow (LOF). The MARS-LMR code is used to analyze the thermal-hydraulic behavior and the safety of reactor. The primary and secondary trip variables are identified and delayed effects were analyzed for each event. The results show that the reactor can be safely cooled down in both primary and secondary trip scenarios for all events, with the secondary trip parameters providing adequate protection. The cumulative damage fraction (CDF) values for the fuel cladding integrity remain within safety acceptance criteria. The study demonstrates the selected trip parameters in RPS is proper to ensure the safety of SALUS for the representative AOOs.
{"title":"Investigation on diversity of trip parameters in reactor protection system for SALUS under various anticipated operational occurrences","authors":"Ji-Woong Han, Sun Rock Choi, In Sub Jun, Seungjoon Baik, Huee-Youl Ye, Jewhan Lee","doi":"10.1016/j.nucengdes.2025.114727","DOIUrl":"10.1016/j.nucengdes.2025.114727","url":null,"abstract":"<div><div>This paper investigates the diversity of trip parameters in the reactor protection system (RPS) of SALUS, a Small, Advanced, Long-cycled, and Ultimate safe Sodium-cooled fast reactor, for various anticipated operational occurrences (AOOs). Three representative AOO events are selected: transient overpower (TOP), loss of heat sink (LOHS), and loss of core flow (LOF). The MARS-LMR code is used to analyze the thermal-hydraulic behavior and the safety of reactor. The primary and secondary trip variables are identified and delayed effects were analyzed for each event. The results show that the reactor can be safely cooled down in both primary and secondary trip scenarios for all events, with the secondary trip parameters providing adequate protection. The cumulative damage fraction (CDF) values for the fuel cladding integrity remain within safety acceptance criteria. The study demonstrates the selected trip parameters in RPS is proper to ensure the safety of SALUS for the representative AOOs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114727"},"PeriodicalIF":2.1,"publicationDate":"2026-01-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145950225","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-10DOI: 10.1016/j.nucengdes.2026.114766
Palash K. Bhowmik, Mauricio E. Tano, SuJong Yoon, Changhu Xing, Silvino A.B. Prieto, Alexander L. Swearingen, Ann M. Phillips, Piyush Sabharwall, Jeffrey J. Giglio
This study covers the onset of flow instability (OFI) preliminary results obtained from leveraging correlations, in addition to the preliminary thermal hydraulics results such as pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment (DDE) for the Massachusetts Institute of Technology Reactor (MITR). Current computational fluid dynamics (CFD) models in fluid structure interaction (FSI) have added the capability of assessing margins to onset of nucleate boiling (ONB). This study initiates the capability to model the margin to OFI and ONB presented for the MITR. Such study is supportive of the United States High Performance Research Reactor (USHPRR) program. Previous studies provided preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the MITR DDE under conservative approximations for plate power distribution. This study focuses on providing insights into the OFI future research direction optimizing the transport of thermal energy, mass-flow rates, flow-channel geometries, and boundary conditions.
{"title":"Margin to onset of nucleate boiling and flow instability studies for preliminary MITR design-demonstration element thermal-hydraulics","authors":"Palash K. Bhowmik, Mauricio E. Tano, SuJong Yoon, Changhu Xing, Silvino A.B. Prieto, Alexander L. Swearingen, Ann M. Phillips, Piyush Sabharwall, Jeffrey J. Giglio","doi":"10.1016/j.nucengdes.2026.114766","DOIUrl":"10.1016/j.nucengdes.2026.114766","url":null,"abstract":"<div><div>This study covers the onset of flow instability (OFI) preliminary results obtained from leveraging correlations, in addition to the preliminary thermal hydraulics results such as pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment (DDE) for the Massachusetts Institute of Technology Reactor (MITR). Current computational fluid dynamics (CFD) models in fluid structure interaction (FSI) have added the capability of assessing margins to onset of nucleate boiling (ONB). This study initiates the capability to model the margin to OFI and ONB presented for the MITR. Such study is supportive of the United States High Performance Research Reactor (USHPRR) program. Previous studies provided preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the MITR DDE under conservative approximations for plate power distribution. This study focuses on providing insights into the OFI future research direction optimizing the transport of thermal energy, mass-flow rates, flow-channel geometries, and boundary conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114766"},"PeriodicalIF":2.1,"publicationDate":"2026-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145978183","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-10DOI: 10.1016/j.nucengdes.2025.114700
Ali Ahmadi , Naser Khaji , Hamid Sadegh-Azar
In this paper, a resonance-focused Critical Excitation (CE) overlay is developed as a secondary check for base-isolated nuclear power plants. The goal is to capture worst-case, yet physically plausible, motions that align with the isolation period and key equipment periods, while remaining compatible with ASCE 4–16 and ASCE 43–19 practices. The method generates three CE bounds (lower, mean, upper) under explicit Arias intensity, PGA, and PGV constraints. At the same time, routine suite-mean results are read against these bounds to flag under-targeting, proximity, or design-extension concerns. Well-established isolated-reactor models, including two-degree-of-freedom with low-damping rubber, lead rubber bearing, friction pendulum system, and six-degree-of-freedom configurations, are used. As a case study, ground motions are selected for the Diablo Canyon site and matched to the SDC-5 Design Response Spectrum. A code-consistent suite of ground motion records forms the design baseline for averaging. A screening study compares acceleration- and displacement-targeted CE objectives. The displacement objective produces higher peaks in isolator displacement, isolation-plane base shear, and floor acceleration for six of seven seed ground motions; therefore, it is adopted for design-level evaluation. With the displacement objective, suite-mean responses are placed within the CE three-bound for key metrics, indicating conservative and stable estimates without missing resonance. The overlay provides clear decision triggers: (a) below the CE lower bound, supplementation or retuning is indicated; (b) near the CE mean, demand capture is adequate and remaining margins are checked; and (c) trending toward the CE upper bound signals a Beyond-Design-Basis Earthquake (BDBE) condition and prompts targeted checks. The CE overlay thus serves as a transparent, code-compatible safety gate and supports BDBE reasoning without any arbitrary multipliers.
{"title":"Beyond-design-basis screening by a three-bound critical excitation envelope for base-isolated nuclear power plants","authors":"Ali Ahmadi , Naser Khaji , Hamid Sadegh-Azar","doi":"10.1016/j.nucengdes.2025.114700","DOIUrl":"10.1016/j.nucengdes.2025.114700","url":null,"abstract":"<div><div>In this paper, a resonance-focused Critical Excitation (CE) overlay is developed as a secondary check for base-isolated nuclear power plants. The goal is to capture worst-case, yet physically plausible, motions that align with the isolation period and key equipment periods, while remaining compatible with ASCE 4–16 and ASCE 43–19 practices. The method generates three CE bounds (lower, mean, upper) under explicit Arias intensity, PGA, and PGV constraints. At the same time, routine suite-mean results are read against these bounds to flag under-targeting, proximity, or design-extension concerns. Well-established isolated-reactor models, including two-degree-of-freedom with low-damping rubber, lead rubber bearing, friction pendulum system, and six-degree-of-freedom configurations, are used. As a case study, ground motions are selected for the Diablo Canyon site and matched to the SDC-5 Design Response Spectrum. A code-consistent suite of ground motion records forms the design baseline for averaging. A screening study compares acceleration- and displacement-targeted CE objectives. The displacement objective produces higher peaks in isolator displacement, isolation-plane base shear, and floor acceleration for six of seven seed ground motions; therefore, it is adopted for design-level evaluation. With the displacement objective, suite-mean responses are placed within the CE three-bound for key metrics, indicating conservative and stable estimates without missing resonance. The overlay provides clear decision triggers: (a) below the CE lower bound, supplementation or retuning is indicated; (b) near the CE mean, demand capture is adequate and remaining margins are checked; and (c) trending toward the CE upper bound signals a Beyond-Design-Basis Earthquake (BDBE) condition and prompts targeted checks. The CE overlay thus serves as a transparent, code-compatible safety gate and supports BDBE reasoning without any arbitrary multipliers.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114700"},"PeriodicalIF":2.1,"publicationDate":"2026-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939545","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-10DOI: 10.1016/j.nucengdes.2025.114737
Aaron Epiney , Gerhard Strydom , Robert Kile , Jonathan Barthle , Izabela Gutowska , Benjamin Nakhnikian-Weintraub , Thanh Hua , Ling Zou , Jun Fang , Krishna Podila , Xianmin Huang , Qi Chen , Tariq Jafri , Geoffrey Waddington , Peter Pfeiffer
Accurate modeling and simulation tools for thermal-hydraulics calculations are a key element needed to design and license new advanced reactors including Small Modular Reactors (SMR) and Microreactors. Uncertainties in modeling and simulation can have significant safety and economic implications.
The High Temperature Test Facility (HTTF) at Oregon State University (OSU) is a scaled integral effects experiment designed to investigate transient behavior in high-temperature gas-cooled prismatic-block nuclear reactors. High-quality measurement data is available from the HTTF that is suitable for a thermal-hydraulics code validation benchmark for gas-cooled reactor simulations.
This paper summarizes individual HTTF modeling efforts to date for tool validation at Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Oregon State University (OSU) and Canadian Nuclear Laboratories (CNL) using system thermal-hydraulics codes, Computational Fluid Dynamics (CFD) codes and system-CFD code couplings. Also, the paper introduces the ongoing OECD Nuclear Energy Agency (NEA) High Temperature Gas Reactor Thermal-Hydraulics (HTGR T/H) benchmark that allows for better comparisons of results between different international modeling teams. The benchmark provides well defined computational problems that include code-to-code comparisons and comparisons to measured data. These problems provide an avenue for quantifying accuracy and identifying sources of uncertainty in thermal-hydraulics calculations, including in measured thermophysical properties, as part of validation for gas-cooled reactor simulation tools.
{"title":"SMR safety through HTTF modeling and benchmark efforts for code validation for gas-cooled reactor applications","authors":"Aaron Epiney , Gerhard Strydom , Robert Kile , Jonathan Barthle , Izabela Gutowska , Benjamin Nakhnikian-Weintraub , Thanh Hua , Ling Zou , Jun Fang , Krishna Podila , Xianmin Huang , Qi Chen , Tariq Jafri , Geoffrey Waddington , Peter Pfeiffer","doi":"10.1016/j.nucengdes.2025.114737","DOIUrl":"10.1016/j.nucengdes.2025.114737","url":null,"abstract":"<div><div>Accurate modeling and simulation tools for thermal-hydraulics calculations are a key element needed to design and license new advanced reactors including Small Modular Reactors (SMR) and Microreactors. Uncertainties in modeling and simulation can have significant safety and economic implications.</div><div>The High Temperature Test Facility (HTTF) at Oregon State University (OSU) is a scaled integral effects experiment designed to investigate transient behavior in high-temperature gas-cooled prismatic-block nuclear reactors. High-quality measurement data is available from the HTTF that is suitable for a thermal-hydraulics code validation benchmark for gas-cooled reactor simulations.</div><div>This paper summarizes individual HTTF modeling efforts to date for tool validation at Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Oregon State University (OSU) and Canadian Nuclear Laboratories (CNL) using system thermal-hydraulics codes, Computational Fluid Dynamics (CFD) codes and system-CFD code couplings. Also, the paper introduces the ongoing OECD Nuclear Energy Agency (NEA) High Temperature Gas Reactor Thermal-Hydraulics (HTGR T/H) benchmark that allows for better comparisons of results between different international modeling teams. The benchmark provides well defined computational problems that include code-to-code comparisons and comparisons to measured data. These problems provide an avenue for quantifying accuracy and identifying sources of uncertainty in thermal-hydraulics calculations, including in measured thermophysical properties, as part of validation for gas-cooled reactor simulation tools.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114737"},"PeriodicalIF":2.1,"publicationDate":"2026-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939574","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-09DOI: 10.1016/j.nucengdes.2026.114752
Alan Matias Avelar , Jian Su , Claudia Giovedi , Fabio de Camargo , Joseph T. Klamo , Oleg Yakimenko
The licensing of new nuclear reactors after the Fukushima accident presents significant challenges due to the complexity of nuclear systems and the stringent regulatory requirements involved. Model-based systems engineering (MBSE) has emerged as a useful approach for managing the development of such complex systems, while Best Estimate Plus Uncertainty (BEPU) methodologies have proven valuable within regulatory frameworks for safety evaluation. However, digital models and databases that are needed to provide evidence that the system meets the specified requirements are usually isolated in discipline-specific data repositories. To address this challenge, this article proposes a model breakdown structure (MBS) methodology, using a set of interconnectable models to seamlessly integrate MBSE, computer-aided engineering (CAE) models, and BEPU analysis. The Brazilian Multipurpose Reactor (RMB) served as the system of interest to exemplify the effectiveness of the proposed methodology. A requirement specification was linked to a finite element analysis (FEA) that estimates the peak cladding temperature in a slow loss of flow accident scenario. Additionally, key design factors are identified using design of experiments (DOE) and analysis of variance (ANOVA). Lastly, Wilks' theorem and Monte Carlo simulations are applied for uncertainty quantification. The results indicate that the 95/95 upper tolerance limit of the peak cladding temperature remains below the onset of nucleate boiling. Furthermore, the utilization of Wilks' theorem can reduce computational cost for uncertainty quantification, and the effect of sampling methods is negligible in Monte Carlo simulations with large sample sizes. This approach can enhance the verification and validation (V&V) of regulatory requirements in the licensing process of new reactors.
{"title":"Integrating best estimate plus uncertainty analysis into model-based systems engineering","authors":"Alan Matias Avelar , Jian Su , Claudia Giovedi , Fabio de Camargo , Joseph T. Klamo , Oleg Yakimenko","doi":"10.1016/j.nucengdes.2026.114752","DOIUrl":"10.1016/j.nucengdes.2026.114752","url":null,"abstract":"<div><div>The licensing of new nuclear reactors after the Fukushima accident presents significant challenges due to the complexity of nuclear systems and the stringent regulatory requirements involved. Model-based systems engineering (MBSE) has emerged as a useful approach for managing the development of such complex systems, while Best Estimate Plus Uncertainty (BEPU) methodologies have proven valuable within regulatory frameworks for safety evaluation. However, digital models and databases that are needed to provide evidence that the system meets the specified requirements are usually isolated in discipline-specific data repositories. To address this challenge, this article proposes a model breakdown structure (MBS) methodology, using a set of interconnectable models to seamlessly integrate MBSE, computer-aided engineering (CAE) models, and BEPU analysis. The Brazilian Multipurpose Reactor (RMB) served as the system of interest to exemplify the effectiveness of the proposed methodology. A requirement specification was linked to a finite element analysis (FEA) that estimates the peak cladding temperature in a slow loss of flow accident scenario. Additionally, key design factors are identified using design of experiments (DOE) and analysis of variance (ANOVA). Lastly, Wilks' theorem and Monte Carlo simulations are applied for uncertainty quantification. The results indicate that the 95/95 upper tolerance limit of the peak cladding temperature remains below the onset of nucleate boiling. Furthermore, the utilization of Wilks' theorem can reduce computational cost for uncertainty quantification, and the effect of sampling methods is negligible in Monte Carlo simulations with large sample sizes. This approach can enhance the verification and validation (V&V) of regulatory requirements in the licensing process of new reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114752"},"PeriodicalIF":2.1,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939640","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-09DOI: 10.1016/j.nucengdes.2025.114748
Graham Macpherson , David Munn , Alan Johnson , Landon Brockmeyer , Elia Merzari , Jerome Solberg
The OECD NEA CSNI WGAMA CFD Task Group ran a benchmark in 2020 and 2021 to assess the predictive capabilities of coupled fluid structure interaction (FSI) CFD analysis methods. This paper presents the predictions made for the open phase of the benchmark using URANS and LES turbulence modelling approaches, and a comparison of the results to the experimental data.
The benchmark comprised a channel containing two inline cylinders in cross-flow. The cylinders were fixed at one end, free at the other, and had measured resonant frequencies and damping properties. The URANS modelling used ANSYS Fluent 2-way coupled to ANSYS Mechanical. The LES modelling used Nek5000, 1-way coupled to Diablo. Comparisons with cross-channel velocity profiles are presented, both for the mean flow and its RMS. Comparisons are also made to the frequency spectra for point measurements of fluid velocity and pressure, and for the accelerations of the free end of each cylinder.
URANS predicts the average velocity profiles relatively well, and is able to predict the velocity and acceleration spectra at the shedding frequency. However, the frequency content at the 4th harmonic of the shedding frequency is low in the URANS flow fields, and so does not excite accelerations at the resonant frequency of the cylinders. LES makes better predictions of the average profiles, and the velocity spectra agree well at both the shedding frequency and at higher frequencies. The 1-way coupled LES results show good agreement for acceleration spectra.
{"title":"Comparison of URANS and LES predictions for the open phase of the OECD NEA CSNI fluid structure interaction CFD benchmark","authors":"Graham Macpherson , David Munn , Alan Johnson , Landon Brockmeyer , Elia Merzari , Jerome Solberg","doi":"10.1016/j.nucengdes.2025.114748","DOIUrl":"10.1016/j.nucengdes.2025.114748","url":null,"abstract":"<div><div>The OECD NEA CSNI WGAMA CFD Task Group ran a benchmark in 2020 and 2021 to assess the predictive capabilities of coupled fluid structure interaction (FSI) CFD analysis methods. This paper presents the predictions made for the open phase of the benchmark using URANS and LES turbulence modelling approaches, and a comparison of the results to the experimental data.</div><div>The benchmark comprised a channel containing two inline cylinders in cross-flow. The cylinders were fixed at one end, free at the other, and had measured resonant frequencies and damping properties. The URANS modelling used ANSYS Fluent 2-way coupled to ANSYS Mechanical. The LES modelling used Nek5000, 1-way coupled to Diablo. Comparisons with cross-channel velocity profiles are presented, both for the mean flow and its RMS. Comparisons are also made to the frequency spectra for point measurements of fluid velocity and pressure, and for the accelerations of the free end of each cylinder.</div><div>URANS predicts the average velocity profiles relatively well, and is able to predict the velocity and acceleration spectra at the shedding frequency. However, the frequency content at the 4th harmonic of the shedding frequency is low in the URANS flow fields, and so does not excite accelerations at the resonant frequency of the cylinders. LES makes better predictions of the average profiles, and the velocity spectra agree well at both the shedding frequency and at higher frequencies. The 1-way coupled LES results show good agreement for acceleration spectra.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114748"},"PeriodicalIF":2.1,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939639","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-08DOI: 10.1016/j.nucengdes.2026.114754
An Ping , Shen Jie , Liu Rui , Liu Dong , He Xiao Qiang , Liu Ting , Hu Ming
Many fixed-point problems are involved in the numerical calculation of reactor high-fidelity simulation. The Gauss-Seidel and the SOR (Successive Over-Relaxation) algorithm are commonly used for iterative acceleration. In this paper, based on Anderson's idea, an acceleration algorithm, known as RCA (Reactor Coupling Calculation Acceleration), is proposed. This algorithm is suitable for multi-disciplinary coupling calculation fixed point iteration for reactors. The RCA algorithm takes the subspace as the iteration object, and the numerical format is determined by minimizing the weighted residual 2-norm, which has good convergence. This paper adopts the RCA algorithm to accelerate the coupled calculation of neutronics and thermal-hydraulics, and calculates the boron critical searching under normal or accident operating conditions. The results show that the RCA algorithm can reduce the number of iterations and improve the calculation efficiency with the same calculation accuracy. Our research provides support for the optimization of existing software and the development of new software.
{"title":"The application of the fixed-point iteration acceleration in the neutronics and thermal-hydraulics coupling calculation","authors":"An Ping , Shen Jie , Liu Rui , Liu Dong , He Xiao Qiang , Liu Ting , Hu Ming","doi":"10.1016/j.nucengdes.2026.114754","DOIUrl":"10.1016/j.nucengdes.2026.114754","url":null,"abstract":"<div><div>Many fixed-point problems are involved in the numerical calculation of reactor high-fidelity simulation. The Gauss-Seidel and the SOR (Successive Over-Relaxation) algorithm are commonly used for iterative acceleration. In this paper, based on Anderson's idea, an acceleration algorithm, known as RCA (Reactor Coupling Calculation Acceleration), is proposed. This algorithm is suitable for multi-disciplinary coupling calculation fixed point iteration for reactors. The RCA algorithm takes the subspace as the iteration object, and the numerical format is determined by minimizing the weighted residual 2-norm, which has good convergence. This paper adopts the RCA algorithm to accelerate the coupled calculation of neutronics and thermal-hydraulics, and calculates the boron critical searching under normal or accident operating conditions. The results show that the RCA algorithm can reduce the number of iterations and improve the calculation efficiency with the same calculation accuracy. Our research provides support for the optimization of existing software and the development of new software.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114754"},"PeriodicalIF":2.1,"publicationDate":"2026-01-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939637","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
{"title":"Corrigendum to “Effects of nanosilica and aggregate type on the mechanical, fracture and shielding features of heavyweight concrete” [Nucl. Eng. Des. 431 (2025) 113713]","authors":"Mohsen Ghorbani , Morteza Biklaryan , Morteza Hosseinali Beygi , Omid Lotfi-Omran","doi":"10.1016/j.nucengdes.2026.114751","DOIUrl":"10.1016/j.nucengdes.2026.114751","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114751"},"PeriodicalIF":2.1,"publicationDate":"2026-01-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145978181","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-08DOI: 10.1016/j.nucengdes.2025.114714
Hernán Ariel Castro , Raul Ariel Rodriguez , Hugo Luis Bianchi
Treatment and conditioning of spent ion exchange resins (IERs) from nuclear facilities is a complex process. The direct immobilization of these materials with a hydraulic binder is usually a first option. However, even the operational procedure of immobilization with cement is not complicated, the volume of final solidified waste form increased significantly and its long-term integrity presents certain limitations.
A strategy internationally considered is to apply a prior treatment step to the spent IERs, mainly thermal treatments, in order to reduce the waste volume and stabilize the product.
In the last few years, our research group has developed a novel technology based on low-temperature thermal treatment of the IERs with steam followed by a High Performance Plasma Treatment (HPPT) of the generated off-gas. The process is capable of achieving high volume reduction factors and a non-reactive solid product.
In the present work, the quality of the solid product obtained in a test bench scale of the process is studied, emphasizing the product compatibility with cement. The solid product embedding with ordinary Portland cement (OPC), without any chemical additives or supplementary materials, was examined. The waste incorporation rate was up to roughly 90% (in volume). The waste form obtained was homogenous and presented compressive strength values around 18 MPa. No evidence of deterioration was observed after 90 days of water immersion.
{"title":"Thermal treatment and high performance plasma treatment applied to spent ion exchange resins: study of solid product embedding with ordinary Portland cement","authors":"Hernán Ariel Castro , Raul Ariel Rodriguez , Hugo Luis Bianchi","doi":"10.1016/j.nucengdes.2025.114714","DOIUrl":"10.1016/j.nucengdes.2025.114714","url":null,"abstract":"<div><div>Treatment and conditioning of spent ion exchange resins (IERs) from nuclear facilities is a complex process. The direct immobilization of these materials with a hydraulic binder is usually a first option. However, even the operational procedure of immobilization with cement is not complicated, the volume of final solidified waste form increased significantly and its long-term integrity presents certain limitations.</div><div>A strategy internationally considered is to apply a prior treatment step to the spent IERs, mainly thermal treatments, in order to reduce the waste volume and stabilize the product.</div><div>In the last few years, our research group has developed a novel technology based on low-temperature thermal treatment of the IERs with steam followed by a High Performance Plasma Treatment (HPPT) of the generated off-gas. The process is capable of achieving high volume reduction factors and a non-reactive solid product.</div><div>In the present work, the quality of the solid product obtained in a test bench scale of the process is studied, emphasizing the product compatibility with cement. The solid product embedding with ordinary Portland cement (OPC), without any chemical additives or supplementary materials, was examined. The waste incorporation rate was up to roughly 90% (in volume). The waste form obtained was homogenous and presented compressive strength values around 18 MPa. No evidence of deterioration was observed after 90 days of water immersion.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114714"},"PeriodicalIF":2.1,"publicationDate":"2026-01-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939638","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The in-house coupled thermal-hydraulic and neutronic code, MTRDYN, has been developed with a three-dimensional capability to solve few-group neutron diffusion equations and thermal-hydraulic parameters for plate type fueled research reactor. The multi-group neutron diffusion equations are addressed through neutron flux factorization within an adiabatic kinetic equation. Heat conduction in the fuel element was computed using the finite difference method, with the heat transfer restricted to the radial direction approximation. This study aims to evaluate the accuracy of the MTRDYN in calculating the behavior of RSG-GAS reactor during steady-state operation. The calculated core parameters include excess reactivity, power peaking factor (PPF), fuel cladding temperature, and coolant temperatures. The coolant and cladding temperature obtained from MTRDYN were validated against measured data from instrumented fuel elements (IFE) located at various positions within the core. The calculated excess reactivity for the first and sixth cores differed from experimental results by −160 pcm and 20.0 pcm, respectively. The total control rod reactivity showed a maximum error of 3.9 % compared to experimental results. No significant differences in kinetic parameters were found compared to the RSG-GAS safety analysis report (SAR). The calculated fuel cladding temperatures showed a maximum deviation of 5.78 %. Based on these calculations, the MTRDYN code demonstrates sufficient accuracy in determining the steady-state neutronic and thermal-hydraulic parameters of the RSG-GAS reactor.
{"title":"The verification and validation of the coupled neutronics thermal-hydraulics code, MTRDYN, for steady-state condition of RSG-GAS reactor","authors":"Surian Pinem , Farisy Yogatama Sulistyo , Peng Hong Liem , Sukmanto Dibyo , Wahid Luthfi","doi":"10.1016/j.nucengdes.2025.114746","DOIUrl":"10.1016/j.nucengdes.2025.114746","url":null,"abstract":"<div><div>The in-house coupled thermal-hydraulic and neutronic code, MTRDYN, has been developed with a three-dimensional capability to solve few-group neutron diffusion equations and thermal-hydraulic parameters for plate type fueled research reactor. The multi-group neutron diffusion equations are addressed through neutron flux factorization within an adiabatic kinetic equation. Heat conduction in the fuel element was computed using the finite difference method, with the heat transfer restricted to the radial direction approximation. This study aims to evaluate the accuracy of the MTRDYN in calculating the behavior of RSG-GAS reactor during steady-state operation. The calculated core parameters include excess reactivity, power peaking factor (PPF), fuel cladding temperature, and coolant temperatures. The coolant and cladding temperature obtained from MTRDYN were validated against measured data from instrumented fuel elements (IFE) located at various positions within the core. The calculated excess reactivity for the first and sixth cores differed from experimental results by −160 pcm and 20.0 pcm, respectively. The total control rod reactivity showed a maximum error of 3.9 % compared to experimental results. No significant differences in kinetic parameters were found compared to the RSG-GAS safety analysis report (SAR). The calculated fuel cladding temperatures showed a maximum deviation of 5.78 %. Based on these calculations, the MTRDYN code demonstrates sufficient accuracy in determining the steady-state neutronic and thermal-hydraulic parameters of the RSG-GAS reactor.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114746"},"PeriodicalIF":2.1,"publicationDate":"2026-01-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939636","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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