Pub Date : 2026-01-14DOI: 10.1016/j.nucengdes.2026.114755
Jeehee Lee , Seong-Su Jeon , Ju-Yeop Park , Hyoung Kyu Cho
The purpose of this study is to develop a robustness assessment methodology for performance evaluation considering the performance characteristics of passive safety systems being introduced in light water reactors and to propose safety analysis guidelines for passive safety systems by evaluating their impact on various performance degradation factors. To develop the methodology, the concerns with the introduction of passive safety systems and the current technical standards for passive safety systems from regulatory bodies around the world were analyzed. Since passive safety systems have less existing operating experience, there is uncertainty about the performance of the system, and it is necessary to prove the applicability of existing system analysis codes. In addition, since a passive safety system does not use devices such as pumps, it is more likely than a conventional safety system that the performance of the system will be degraded by changes in the internal or external environment. Therefore, this study developed a robustness assessment methodology consisting of seven steps to evaluate the impact of issues on the introduction of a passive safety system and to demonstrate the ability of the passive system to perform safety functions.
{"title":"Development of robustness assessment methodology for passive safety system against potential performance issue","authors":"Jeehee Lee , Seong-Su Jeon , Ju-Yeop Park , Hyoung Kyu Cho","doi":"10.1016/j.nucengdes.2026.114755","DOIUrl":"10.1016/j.nucengdes.2026.114755","url":null,"abstract":"<div><div>The purpose of this study is to develop a robustness assessment methodology for performance evaluation considering the performance characteristics of passive safety systems being introduced in light water reactors and to propose safety analysis guidelines for passive safety systems by evaluating their impact on various performance degradation factors. To develop the methodology, the concerns with the introduction of passive safety systems and the current technical standards for passive safety systems from regulatory bodies around the world were analyzed. Since passive safety systems have less existing operating experience, there is uncertainty about the performance of the system, and it is necessary to prove the applicability of existing system analysis codes. In addition, since a passive safety system does not use devices such as pumps, it is more likely than a conventional safety system that the performance of the system will be degraded by changes in the internal or external environment. Therefore, this study developed a robustness assessment methodology consisting of seven steps to evaluate the impact of issues on the introduction of a passive safety system and to demonstrate the ability of the passive system to perform safety functions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114755"},"PeriodicalIF":2.1,"publicationDate":"2026-01-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981790","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 : 2026-01-14DOI: 10.1016/j.nucengdes.2026.114762
Yohan Kim, Hyungdae Kim
The bent heat pipe, which connects a compact reactor core to a relatively large power conversion system, is a critical component in the design of kilowatt-scale heat pipe-cooled reactors for space nuclear applications. Conventional screen wicks (SWs) used in straight heat pipes are prone to structural damage when bent, resulting in a loss of capillary performance. To address this issue, bendable braided-wire wicks (BWWs) have recently been proposed as a promising alternative due to their ability to maintain capillary functionality even under bending conditions. However, comprehensive studies on the capillary performance of BWWs in bent configurations remain limited. This study experimentally investigates the capillary flow characteristics of BWWs in both straight and bent geometries. The porosity was calculated through a geometric analysis of a unit cell in the wick and found to be 0.215. Capillary rise in a vertically oriented straight wick was visualized and quantitatively assessed using infrared thermography. The effective pore radius and permeability were determined by fitting the experimental data of the sample, yielding values were 67.3 μm and m2, respectively. Subsequently, a series of rate-of-rise experiments was conducted on bent BWWs to evaluate the impact of bending on their capillary behavior. The experimental results demonstrated that the BWW maintains its intrinsic capillary properties irrespective of geometric configuration. Finally, theoretical models were proposed to predict the porosity, permeability, and effective pore radius of the BWW structure. The predicted values agreed with the experimental measurements within a margin of approximately 20%.
{"title":"Characterization of braided-wire wicks for bent heat pipe applications: Experiments and modeling","authors":"Yohan Kim, Hyungdae Kim","doi":"10.1016/j.nucengdes.2026.114762","DOIUrl":"10.1016/j.nucengdes.2026.114762","url":null,"abstract":"<div><div>The bent heat pipe, which connects a compact reactor core to a relatively large power conversion system, is a critical component in the design of kilowatt-scale heat pipe-cooled reactors for space nuclear applications. Conventional screen wicks (SWs) used in straight heat pipes are prone to structural damage when bent, resulting in a loss of capillary performance. To address this issue, bendable braided-wire wicks (BWWs) have recently been proposed as a promising alternative due to their ability to maintain capillary functionality even under bending conditions. However, comprehensive studies on the capillary performance of BWWs in bent configurations remain limited. This study experimentally investigates the capillary flow characteristics of BWWs in both straight and bent geometries. The porosity was calculated through a geometric analysis of a unit cell in the wick and found to be 0.215. Capillary rise in a vertically oriented straight wick was visualized and quantitatively assessed using infrared thermography. The effective pore radius and permeability were determined by fitting the experimental data of the sample, yielding values were 67.3 μm and <span><math><mn>1.8</mn><mo>×</mo><msup><mn>10</mn><mrow><mo>−</mo><mn>11</mn></mrow></msup></math></span> m<sup>2</sup>, respectively. Subsequently, a series of rate-of-rise experiments was conducted on bent BWWs to evaluate the impact of bending on their capillary behavior. The experimental results demonstrated that the BWW maintains its intrinsic capillary properties irrespective of geometric configuration. Finally, theoretical models were proposed to predict the porosity, permeability, and effective pore radius of the BWW structure. The predicted values agreed with the experimental measurements within a margin of approximately 20%.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114762"},"PeriodicalIF":2.1,"publicationDate":"2026-01-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981789","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 : 2026-01-13DOI: 10.1016/j.nucengdes.2026.114756
Rui Hou , Shaowei Tang , Jingliang Zhang , Yi Lei , Bin Zhang , Jianqiang Shan
Under Loss-of-Flow (LOF) accident conditions, the boiling behavior of the coolant has a decisive impact on core integrity. A multi-bubble slug model has been implemented and integrated within the integrated severe accident analysis code ISAA-Na to dynamically simulate the two-phase flow and heat transfer behavior in sodium-cooled fast reactors (SFRs) under LOF conditions. To validate the model's effectiveness, experiments BI1, E8, and EFM1 from the French CABRI facility were selected as benchmarks. The predictive capability of the model was assessed through a systematic comparison of the calculation results from ISAA-Na with experimental data and published results from other mainstream codes. The assessment indicates that the model accurately captures key physical phenomena, at boiling inception, the saturation temperature is predicted with a relative error of 0.12%. For the subsequent two-phase conditions, the model captures boiling initiation times with an absolute error of less than 0.3 s and axial locations within a relative error of 6%, while also accurately reproducing interface propagation and flow oscillations. This confirms the reliability of the model's implementation in ISAA-Na for SFR safety analysis, providing a robust basis for predicting subsequent accident progression.
{"title":"Validation of the ISAA-Na two-phase model for sodium-cooled fast reactors: An assessment against CABRI LOF experiments","authors":"Rui Hou , Shaowei Tang , Jingliang Zhang , Yi Lei , Bin Zhang , Jianqiang Shan","doi":"10.1016/j.nucengdes.2026.114756","DOIUrl":"10.1016/j.nucengdes.2026.114756","url":null,"abstract":"<div><div>Under Loss-of-Flow (LOF) accident conditions, the boiling behavior of the coolant has a decisive impact on core integrity. A multi-bubble slug model has been implemented and integrated within the integrated severe accident analysis code ISAA-Na to dynamically simulate the two-phase flow and heat transfer behavior in sodium-cooled fast reactors (SFRs) under LOF conditions. To validate the model's effectiveness, experiments BI1, E8, and EFM1 from the French CABRI facility were selected as benchmarks. The predictive capability of the model was assessed through a systematic comparison of the calculation results from ISAA-Na with experimental data and published results from other mainstream codes. The assessment indicates that the model accurately captures key physical phenomena, at boiling inception, the saturation temperature is predicted with a relative error of 0.12%. For the subsequent two-phase conditions, the model captures boiling initiation times with an absolute error of less than 0.3 s and axial locations within a relative error of 6%, while also accurately reproducing interface propagation and flow oscillations. This confirms the reliability of the model's implementation in ISAA-Na for SFR safety analysis, providing a robust basis for predicting subsequent accident progression.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114756"},"PeriodicalIF":2.1,"publicationDate":"2026-01-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981788","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 : 2026-01-13DOI: 10.1016/j.nucengdes.2026.114763
J. Schäfer , S. Taş , U. Hampel
This work presents a novel training approach for machine learning based instance segmentation, without the need for manual annotated datasets. With applications in experimental investigations in bubbly flows in reactor safety research. This semi-supervised process consists of two different neural networks, a conditional generative adversarial network used for data generation and a U-net style convolutional neural network for instance segmentation. We validated the approach using a fully automated experimental setup creating layered bubble curtains, which enables an evaluation of the bubble size distribution with and without overlapping bubbles. The final neural network is able to measure the average bubble size as well as to recreate the bubble size distribution accurately.
{"title":"Gas bubble detection and segmentation using a machine learning approach leveraging semi-supervised training","authors":"J. Schäfer , S. Taş , U. Hampel","doi":"10.1016/j.nucengdes.2026.114763","DOIUrl":"10.1016/j.nucengdes.2026.114763","url":null,"abstract":"<div><div>This work presents a novel training approach for machine learning based instance segmentation, without the need for manual annotated datasets. With applications in experimental investigations in bubbly flows in reactor safety research. This semi-supervised process consists of two different neural networks, a conditional generative adversarial network used for data generation and a U-net style convolutional neural network for instance segmentation. We validated the approach using a fully automated experimental setup creating layered bubble curtains, which enables an evaluation of the bubble size distribution with and without overlapping bubbles. The final neural network is able to measure the average bubble size as well as to recreate the bubble size distribution accurately.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114763"},"PeriodicalIF":2.1,"publicationDate":"2026-01-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981787","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 : 2026-01-12DOI: 10.1016/j.nucengdes.2026.114769
Björn Engström, Weimin Ma
A simplified neutron-kinetics model for recriticality calculations has been developed and implemented within the severe accident code MELCOR. The model partitions the core into active and inactive regions and determines neutron flux using factorization with adiabatic and prompt-jump approximations. Homogenization of pre-calculated core cell multiplication factors, together with Doppler and xenon reactivity defects calculated by perturbation theory, avoids explicit treatment of cross-sections. This approach allows recriticality analysis to be performed seamlessly within MELCOR at standard time-steps and computational cost. Comparisons with SARA project simulations show reasonable agreement. Integrated with MELCOR, the model extends its capabilities to predict reactivity, recriticality timing, fission power, fuel temperatures, and xenon transients. The model was applied to station blackout scenarios in a Nordic BWR with varying low-pressure safety injection timing and two decay-heat curves. Simulations suggest that the time window for recriticality may be substantial if the core periphery remains intact and coolant flow through a breached vessel is limited. In cases where recriticality occurred, fission power evolution was irregular, and even relatively low fission power accelerated containment heat-up and pressurization. These results demonstrate that the simplified model provides an efficient tool for investigating recriticality phenomena, their impact on severe accident progression, and the effectiveness of mitigation strategies under uncertainty and sensitivity analyses.
{"title":"Development and application of a simplified neutron-kinetics model for severe accident recriticality assessment","authors":"Björn Engström, Weimin Ma","doi":"10.1016/j.nucengdes.2026.114769","DOIUrl":"10.1016/j.nucengdes.2026.114769","url":null,"abstract":"<div><div>A simplified neutron-kinetics model for recriticality calculations has been developed and implemented within the severe accident code MELCOR. The model partitions the core into active and inactive regions and determines neutron flux using factorization with adiabatic and prompt-jump approximations. Homogenization of pre-calculated core cell multiplication factors, together with Doppler and xenon reactivity defects calculated by perturbation theory, avoids explicit treatment of cross-sections. This approach allows recriticality analysis to be performed seamlessly within MELCOR at standard time-steps and computational cost. Comparisons with SARA project simulations show reasonable agreement. Integrated with MELCOR, the model extends its capabilities to predict reactivity, recriticality timing, fission power, fuel temperatures, and xenon transients. The model was applied to station blackout scenarios in a Nordic BWR with varying low-pressure safety injection timing and two decay-heat curves. Simulations suggest that the time window for recriticality may be substantial if the core periphery remains intact and coolant flow through a breached vessel is limited. In cases where recriticality occurred, fission power evolution was irregular, and even relatively low fission power accelerated containment heat-up and pressurization. These results demonstrate that the simplified model provides an efficient tool for investigating recriticality phenomena, their impact on severe accident progression, and the effectiveness of mitigation strategies under uncertainty and sensitivity analyses.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114769"},"PeriodicalIF":2.1,"publicationDate":"2026-01-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145950224","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 : 2026-01-12DOI: 10.1016/j.nucengdes.2025.114696
Quazi Md. Zobaer Shah , Md Mahabub Hasan Mousum , Debashis Datta , Md. Arefin Kowser , Mohammad Asaduzzaman Chowdhury , Md. Shojib Mia , Shahpara Sheikh Dola
Grid-to-rod fretting (GTRF) represents a persistent challenge to the mechanical integrity and service life of nuclear fuel assemblies across various reactor technologies. As a leading cause of fuel failures in pressurized water reactors and a growing concern in other light water reactor systems, GTRF has drawn significant attention from both experimental and computational communities. Recent investigations have expanded in scope and precision, with advances in multiphysics simulation frameworks, in-reactor diagnostics, and surface-engineered cladding materials. These studies have illuminated complex dependencies on flow turbulence, spacer grid geometry, contact dynamics, and material behavior under irradiation, yet several technical uncertainties continue to limit predictive confidence. This work presents a comprehensive synthesis of developments surrounding GTRF, highlighting current approaches that span high-resolution fluid-structure interaction modeling, fretting-wear characterization, and emerging mitigation strategies. Specific phenomena such as fretting wear response under variable coolant conditions, coating-induced changes in contact fatigue, and the evolving mechanical role of spacer supports are examined through a multidisciplinary lens. While no single mechanism dominates across all contexts, the convergence of insights from structural mechanics, tribology, and reactor operation points toward integrated pathways for addressing the problem. GTRF arises from tightly coupled turbulent forcing, nonlinear rod-support dynamics and evolving material states. Integrated LES→FSI→wear pipelines, validated against autoclave and flow-loop experiments, offer the most promising path toward predictive life-assessment. It was recommended that coordinated high-burnup testing, open benchmark datasets, and development of reduced-order multiphysics tools with uncertainty quantification to enable quantified life-predictions for advanced claddings and higher-burnup operation. Viewing GTRF as a systems-level challenge highlights its implications for fuel reliability and safety and underscores the need for focused research to achieve quantitative, licensing-grade predictions.
{"title":"Integrated multiphysics assessment of grid-to-rod fretting mitigation: enhancing nuclear safety through improved fuel integrity","authors":"Quazi Md. Zobaer Shah , Md Mahabub Hasan Mousum , Debashis Datta , Md. Arefin Kowser , Mohammad Asaduzzaman Chowdhury , Md. Shojib Mia , Shahpara Sheikh Dola","doi":"10.1016/j.nucengdes.2025.114696","DOIUrl":"10.1016/j.nucengdes.2025.114696","url":null,"abstract":"<div><div>Grid-to-rod fretting (GTRF) represents a persistent challenge to the mechanical integrity and service life of nuclear fuel assemblies across various reactor technologies. As a leading cause of fuel failures in pressurized water reactors and a growing concern in other light water reactor systems, GTRF has drawn significant attention from both experimental and computational communities. Recent investigations have expanded in scope and precision, with advances in multiphysics simulation frameworks, in-reactor diagnostics, and surface-engineered cladding materials. These studies have illuminated complex dependencies on flow turbulence, spacer grid geometry, contact dynamics, and material behavior under irradiation, yet several technical uncertainties continue to limit predictive confidence. This work presents a comprehensive synthesis of developments surrounding GTRF, highlighting current approaches that span high-resolution fluid-structure interaction modeling, fretting-wear characterization, and emerging mitigation strategies. Specific phenomena such as fretting wear response under variable coolant conditions, coating-induced changes in contact fatigue, and the evolving mechanical role of spacer supports are examined through a multidisciplinary lens. While no single mechanism dominates across all contexts, the convergence of insights from structural mechanics, tribology, and reactor operation points toward integrated pathways for addressing the problem. GTRF arises from tightly coupled turbulent forcing, nonlinear rod-support dynamics and evolving material states. Integrated LES→FSI→wear pipelines, validated against autoclave and flow-loop experiments, offer the most promising path toward predictive life-assessment. It was recommended that coordinated high-burnup testing, open benchmark datasets, and development of reduced-order multiphysics tools with uncertainty quantification to enable quantified life-predictions for advanced claddings and higher-burnup operation. Viewing GTRF as a systems-level challenge highlights its implications for fuel reliability and safety and underscores the need for focused research to achieve quantitative, licensing-grade predictions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114696"},"PeriodicalIF":2.1,"publicationDate":"2026-01-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145950226","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 : 2026-01-12DOI: 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.
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