Accurate burn up calculations are critical for nuclear reactor design, particularly for determining the nuclear concentrations of fuel isotopes and fission products throughout the reactor cycle. An updated cross-sectional library is essential for effective fuel behavior analysis and management. This study aims to develop a tailored cross-sectional library for the VVER-1000 reactor to enhance the accuracy of burn up calculations using the ORIGEN2 code, leveraging the ENDF reference library. The Monte Carlo N-Particle (MCNPX) code was used to generate the required cross-sectionals, which were then integrated into ORIGEN2 for burn up calculations. The results were compared with those obtained using the existing library. The new library demonstrates moderately improved accuracy and computational efficiency for burn up calculations in the VVER-1000 reactor compared to the previous library.
{"title":"Leveraging ENDF data in an enhanced ORIGEN2 library for advanced VVER-1000 fuel management","authors":"Saeedeh Arabzadeh , Seyed Pezhman Shirmardi , Nasser Mansour Shariflou","doi":"10.1016/j.nucengdes.2026.114758","DOIUrl":"10.1016/j.nucengdes.2026.114758","url":null,"abstract":"<div><div>Accurate burn up calculations are critical for nuclear reactor design, particularly for determining the nuclear concentrations of fuel isotopes and fission products throughout the reactor cycle. An updated cross-sectional library is essential for effective fuel behavior analysis and management. This study aims to develop a tailored cross-sectional library for the VVER-1000 reactor to enhance the accuracy of burn up calculations using the ORIGEN2 code, leveraging the ENDF reference library. The Monte Carlo N-Particle (MCNPX) code was used to generate the required cross-sectionals, which were then integrated into ORIGEN2 for burn up calculations. The results were compared with those obtained using the existing library. The new library demonstrates moderately improved accuracy and computational efficiency for burn up calculations in the VVER-1000 reactor compared to the previous library.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114758"},"PeriodicalIF":2.1,"publicationDate":"2026-01-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146057607","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 : 2026-01-28DOI: 10.1016/j.nucengdes.2026.114790
P. Pfahl , B. Kędzierska , I. Kim , A. Chambon , L. Fischer , L. Bureš , J. Groth-Jensen , Y. Kim , A. Rineiski , B. Lauritzen
Verification of nuclear codes is an important step in licensing nuclear reactors. For molten salt reactors, the involved physics phenomena are strongly coupled and include those introduced by the movement of liquid fuel that are not present at nominal conditions in solid fuel reactors. This movement of fuel inside and outside the core poses new simulation challenges. In this paper, a benchmark for a graphite-moderated molten salt reactor with a simplified out-of-core model is proposed and studied. The benchmark addresses both neutronics and thermal-hydraulics phenomena, including the delayed neutron precursor drift inside and outside of the active core region, as well as the temperature feedback. As for the thermal-hydraulics, a laminar flow field with conjugate heat transfer, delayed neutron precursor movement, and a simplified heat exchanger is modeled. The benchmark is investigated with the MOOSE tools Griffin and Squirrel, coupled with the MOOSE internal thermal-hydraulics abilities, the Monte Carlo code iMC coupled with OpenFOAM, Nek5000 with a custom point kinetics solver, the coupled neutronics and fluid dynamics code SIMMER with capabilities for severe accident simulations, and the Modelica-based library TRANSFORM. By employing a variety of high- and low-fidelity modeling approaches, a robust comparison across different codes is ensured. OpenMC and Serpent are employed as reference codes to verify the correct implementation of the neutronics. This paper provides a comprehensive comparison of the strengths and weaknesses of the codes and their underlying modeling assumptions. It examines how modeling assumptions affect the steady-state solution and how they propagate into the transient analysis.
{"title":"Multi-Physics Benchmark for a Thermal Molten Salt Reactor","authors":"P. Pfahl , B. Kędzierska , I. Kim , A. Chambon , L. Fischer , L. Bureš , J. Groth-Jensen , Y. Kim , A. Rineiski , B. Lauritzen","doi":"10.1016/j.nucengdes.2026.114790","DOIUrl":"10.1016/j.nucengdes.2026.114790","url":null,"abstract":"<div><div>Verification of nuclear codes is an important step in licensing nuclear reactors. For molten salt reactors, the involved physics phenomena are strongly coupled and include those introduced by the movement of liquid fuel that are not present at nominal conditions in solid fuel reactors. This movement of fuel inside and outside the core poses new simulation challenges. In this paper, a benchmark for a graphite-moderated molten salt reactor with a simplified out-of-core model is proposed and studied. The benchmark addresses both neutronics and thermal-hydraulics phenomena, including the delayed neutron precursor drift inside and outside of the active core region, as well as the temperature feedback. As for the thermal-hydraulics, a laminar flow field with conjugate heat transfer, delayed neutron precursor movement, and a simplified heat exchanger is modeled. The benchmark is investigated with the MOOSE tools Griffin and Squirrel, coupled with the MOOSE internal thermal-hydraulics abilities, the Monte Carlo code iMC coupled with OpenFOAM, Nek5000 with a custom point kinetics solver, the coupled neutronics and fluid dynamics code SIMMER with capabilities for severe accident simulations, and the Modelica-based library TRANSFORM. By employing a variety of high- and low-fidelity modeling approaches, a robust comparison across different codes is ensured. OpenMC and Serpent are employed as reference codes to verify the correct implementation of the neutronics. This paper provides a comprehensive comparison of the strengths and weaknesses of the codes and their underlying modeling assumptions. It examines how modeling assumptions affect the steady-state solution and how they propagate into the transient analysis.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114790"},"PeriodicalIF":2.1,"publicationDate":"2026-01-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078926","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 : 2026-01-28DOI: 10.1016/j.nucengdes.2026.114792
Jin-Young Park, Tae-Hyun Kwon
Accurate modeling of containment leakage is essential for predicting radioactive release during severe nuclear accidents. Although simplified pressure-based methods provide general estimates, they lack the ability to capture damage progression and determine leakage locations caused by structural failures, such as concrete cracking and liner tearing. This study examines the structural response and leakage behavior of a prestressed concrete containment building representative of the APR-1400 under ultimate internal pressure. A comprehensive three-dimensional nonlinear finite-element (FE) model was developed, and the predicted internal pressures corresponding to characteristic strain levels in major components were consistent with those of the Sandia National Laboratories (SNL) test data and simplified calculations, validating the reliability of the FE model. Four leakage prediction methods were employed to evaluate leakage rates, incorporating both permeation-based and crack development approaches. The analysis reveals that leakage initiates primarily at discontinuities, such as equipment hatch, personnel airlocks, and penetrations, and subsequently propagates to free-field regions as pressure increases. In addition, the relationship between predicted leakage rates and liner plate failure was investigated. Lower liner strain thresholds result in earlier onset and greater magnitude of leakage, emphasizing the critical role of the liner plate in containment integrity. These findings enhance the understanding of leakage mechanisms and provide a robust framework for more accurate integrity assessments of containment buildings. Furthermore, FE-based leakage prediction methods show strong potential for integration into severe accident codes, enabling a more realistic representation of the relationship between containment leakage rate and internal pressure.
{"title":"Prediction of leakage rates from containment buildings under ultimate internal pressure","authors":"Jin-Young Park, Tae-Hyun Kwon","doi":"10.1016/j.nucengdes.2026.114792","DOIUrl":"10.1016/j.nucengdes.2026.114792","url":null,"abstract":"<div><div>Accurate modeling of containment leakage is essential for predicting radioactive release during severe nuclear accidents. Although simplified pressure-based methods provide general estimates, they lack the ability to capture damage progression and determine leakage locations caused by structural failures, such as concrete cracking and liner tearing. This study examines the structural response and leakage behavior of a prestressed concrete containment building representative of the APR-1400 under ultimate internal pressure. A comprehensive three-dimensional nonlinear finite-element (FE) model was developed, and the predicted internal pressures corresponding to characteristic strain levels in major components were consistent with those of the Sandia National Laboratories (SNL) test data and simplified calculations, validating the reliability of the FE model. Four leakage prediction methods were employed to evaluate leakage rates, incorporating both permeation-based and crack development approaches. The analysis reveals that leakage initiates primarily at discontinuities, such as equipment hatch, personnel airlocks, and penetrations, and subsequently propagates to free-field regions as pressure increases. In addition, the relationship between predicted leakage rates and liner plate failure was investigated. Lower liner strain thresholds result in earlier onset and greater magnitude of leakage, emphasizing the critical role of the liner plate in containment integrity. These findings enhance the understanding of leakage mechanisms and provide a robust framework for more accurate integrity assessments of containment buildings. Furthermore, FE-based leakage prediction methods show strong potential for integration into severe accident codes, enabling a more realistic representation of the relationship between containment leakage rate and internal pressure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114792"},"PeriodicalIF":2.1,"publicationDate":"2026-01-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078930","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 : 2026-01-27DOI: 10.1016/j.nucengdes.2026.114791
Hang Jing , Jing Li , Xiaoxi Chen , Qingpei Xiang , Rende Ze , Heng Yan , Liqun Shi , Shuming Peng
Performance and structural optimization of milliwatt-level radioisotope thermoelectric generators (RTGs) with end-mounted thermoelectric modules (TEMs) are investigated. A one-dimensional heat transfer model was developed to analyze temperature distribution and maximum output power (Pmax) of the RTG. The sensitivity of Pmax to TEM length (L) and cross-sectional area (A) was evaluated for RTGs using five thermoelectric materials. Results show that longer L and optimized A enhance the temperature difference (ΔT) and Pmax. For a Bi2Te3-based RTG with single end TEM (RTG-1), optimal Pmax reached 160.19 mW on Earth at L = 28 mm and A = 292.41 mm2, and 273.17 mW on Titan at L = 28 mm and A = 161.29 mm2. Dual-end TEM configurations (RTG-2) yielded identical power outputs. COMSOL simulations validated the model with >90% accuracy. Thermal contact resistance (RC) analysis revealed higher RC necessitates larger L/A ratios for optimal performance. The model provides a versatile tool for designing RTGs with diverse thermoelectric materials.
研究了端装热电模块的毫瓦级放射性同位素热电发生器（rtg）的性能和结构优化。建立了一维传热模型，分析了RTG的温度分布和最大输出功率（Pmax）。利用5种热电材料对rtg进行了Pmax对TEM长度(L)和截面积(A)的敏感性评价。结果表明，较长的L和优化后的A增大了温差（ΔT）和Pmax。对于基于bi2te3的单端TEM RTG (RTG-1)，在L = 28 mm, a = 292.41 mm2时，地球上的最佳Pmax为160.19 mW，在泰坦上的最佳Pmax为273.17 mW, L = 28 mm, a = 161.29 mm2。双端TEM配置（RTG-2）产生相同的功率输出。COMSOL仿真验证了该模型的准确率为90%。热接触电阻（RC）分析表明，较高的RC需要较大的L/A比才能获得最佳性能。该模型为设计具有不同热电材料的rtg提供了一个通用工具。
{"title":"Performance and structural optimization of milliwatt-level radioisotope thermoelectric generators with end-mounted thermoelectric modules","authors":"Hang Jing , Jing Li , Xiaoxi Chen , Qingpei Xiang , Rende Ze , Heng Yan , Liqun Shi , Shuming Peng","doi":"10.1016/j.nucengdes.2026.114791","DOIUrl":"10.1016/j.nucengdes.2026.114791","url":null,"abstract":"<div><div>Performance and structural optimization of milliwatt-level radioisotope thermoelectric generators (RTGs) with end-mounted thermoelectric modules (TEMs) are investigated. A one-dimensional heat transfer model was developed to analyze temperature distribution and maximum output power (<em>P</em><sub><em>max</em></sub>) of the RTG. The sensitivity of <em>P</em><sub><em>max</em></sub> to TEM length <em>(L)</em> and cross-sectional area (<em>A</em>) was evaluated for RTGs using five thermoelectric materials. Results show that longer <em>L</em> and optimized <em>A</em> enhance the temperature difference (<em>ΔT</em>) and <em>P</em><sub><em>max</em></sub>. For a Bi<sub>2</sub>Te<sub>3</sub>-based RTG with single end TEM (RTG-1), optimal <em>P</em><sub><em>max</em></sub> reached 160.19 mW on Earth at <em>L</em> = 28 mm and <em>A</em> = 292.41 mm<sup>2</sup>, and 273.17 mW on Titan at <em>L</em> = 28 mm and <em>A</em> = 161.29 mm<sup>2</sup>. Dual-end TEM configurations (RTG-2) yielded identical power outputs. COMSOL simulations validated the model with >90% accuracy. Thermal contact resistance (<em>R</em><sub><em>C</em></sub>) analysis revealed higher <em>R</em><sub><em>C</em></sub> necessitates larger <em>L</em>/<em>A</em> ratios for optimal performance. The model provides a versatile tool for designing RTGs with diverse thermoelectric materials.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114791"},"PeriodicalIF":2.1,"publicationDate":"2026-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078931","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 : 2026-01-27DOI: 10.1016/j.nucengdes.2026.114760
Hua Rong , Xuan Zhang , Yajing Shen , Shuai Tan , Xinglang Fan , Yang Du , Yifeng Feng
This study investigates the prestressed concrete containment vessel (PCCV) of the Qinshan Phase II nuclear power plant, focusing on the nonlinear thermo-mechanical coupling behavior and probabilistic safety evaluation of its load capacity under accident pressure conditions. An energy-based elastoplastic damage constitutive model for concrete at elevated temperatures was employed to develop a refined finite element model of the PCCV, comprehensively considering the effects of temperature gradients, material degradation, and thermo-mechanical coupling. Numerical simulation reveals the whole process of mechanical behaviour of the containment from elastic response, damage evolution to final failure. The results show that the stress concentration around the equipment opening was caused by the geometric discontinuity, which became the weak link of the structure and the first area of damage. To further quantify the influence of the variability in concrete strength, a stochastic damage response analysis was conducted based on the probability density evolution theory (PDET). The probabilistic evolution of damage and displacement responses at key locations was obtained, and the time-dependent reliability of the structure under different damage thresholds was evaluated. The results indicate that the randomness of concrete material strength significantly affects the damage propagation path and failure mode of the containment structure. The proposed analysis framework provides a theoretical and numerical foundation for risk assessment and reliability-based design of nuclear containment structures under accident conditions.
{"title":"Thermo-mechanical coupled analysis and probabilistic safety evaluation of the prestressed concrete containment vessel under accident pressure","authors":"Hua Rong , Xuan Zhang , Yajing Shen , Shuai Tan , Xinglang Fan , Yang Du , Yifeng Feng","doi":"10.1016/j.nucengdes.2026.114760","DOIUrl":"10.1016/j.nucengdes.2026.114760","url":null,"abstract":"<div><div>This study investigates the prestressed concrete containment vessel (PCCV) of the Qinshan Phase II nuclear power plant, focusing on the nonlinear thermo-mechanical coupling behavior and probabilistic safety evaluation of its load capacity under accident pressure conditions. An energy-based elastoplastic damage constitutive model for concrete at elevated temperatures was employed to develop a refined finite element model of the PCCV, comprehensively considering the effects of temperature gradients, material degradation, and thermo-mechanical coupling. Numerical simulation reveals the whole process of mechanical behaviour of the containment from elastic response, damage evolution to final failure. The results show that the stress concentration around the equipment opening was caused by the geometric discontinuity, which became the weak link of the structure and the first area of damage. To further quantify the influence of the variability in concrete strength, a stochastic damage response analysis was conducted based on the probability density evolution theory (PDET). The probabilistic evolution of damage and displacement responses at key locations was obtained, and the time-dependent reliability of the structure under different damage thresholds was evaluated. The results indicate that the randomness of concrete material strength significantly affects the damage propagation path and failure mode of the containment structure. The proposed analysis framework provides a theoretical and numerical foundation for risk assessment and reliability-based design of nuclear containment structures under accident conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114760"},"PeriodicalIF":2.1,"publicationDate":"2026-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078929","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 : 2026-01-26DOI: 10.1016/j.nucengdes.2026.114757
S. Amirkhosravi , A. Scolaro , F. van Niekerk , M.H. du Toit , A. Pautz
MSRs stand out as prominent candidates among advanced reactor designs, addressing the global demand for safer and more sustainable nuclear energy. Accurate multi-physics modelling is essential for the advancement of MSR technology, particularly for understanding the thermos-hydraulic behaviour of graphite under irradiation. This study focuses on developing and implementing a high-fidelity methodology within the GeN-Foam code to model graphite temperature distribution within porous-medium multi-physics simulations, using the MSRE as a benchmark. The approach combines thermal-hydraulic and neutronic modelling by using Serpent-generated cross-section data as inputs for the Gen-Foam neutronic solver. Validation against MSRE measurements performed at ORNL benchmark data confirms the framework's reliability. The axial temperature distribution yields a Mean Absolute Percentage Error (MAPE) of 1.09%, while the radial distribution shows a MAPE of 0.62%. The average graphite temperature of 935.6 K is consistent with the ORNL reference value of 936.4 K under steady-state conditions.
{"title":"Modelling of MSRE graphite temperature in porous-medium multi-physics simulations","authors":"S. Amirkhosravi , A. Scolaro , F. van Niekerk , M.H. du Toit , A. Pautz","doi":"10.1016/j.nucengdes.2026.114757","DOIUrl":"10.1016/j.nucengdes.2026.114757","url":null,"abstract":"<div><div>MSRs stand out as prominent candidates among advanced reactor designs, addressing the global demand for safer and more sustainable nuclear energy. Accurate multi-physics modelling is essential for the advancement of MSR technology, particularly for understanding the thermos-hydraulic behaviour of graphite under irradiation. This study focuses on developing and implementing a high-fidelity methodology within the GeN-Foam code to model graphite temperature distribution within porous-medium multi-physics simulations, using the MSRE as a benchmark. The approach combines thermal-hydraulic and neutronic modelling by using Serpent-generated cross-section data as inputs for the <strong>Gen-Foam</strong> neutronic solver. Validation against MSRE measurements performed at ORNL benchmark data confirms the framework's reliability. The axial temperature distribution yields a Mean Absolute Percentage Error (MAPE) of 1.09%, while the radial distribution shows a MAPE of 0.62%. The average graphite temperature of 935.6 K is consistent with the ORNL reference value of 936.4 K under steady-state conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114757"},"PeriodicalIF":2.1,"publicationDate":"2026-01-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078927","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 : 2026-01-25DOI: 10.1016/j.nucengdes.2026.114789
K. Khumsa-Ang , M. Fulger , A. Sáez Maderuelo , A. Toivonen , M. Sipova , D. Marusakova , L. Zhang , J. Macák , R. Novotny
This document presents a summary of the most relevant research and development (R&D) carried out to support the development of the only generation IV water-cooled reactor endorsed by the Generation IV International Forum (GIF).The coolant of the proposed reactor operates at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of high-grade process heat. Several collaborations have been established to support this technology under the GIF umbrella as well as through other international avenues; as a result, the development work is bolstered by a collective effort between numerous R&D institutions across Asia, Europe, and North America. The Joint European Canadian Chinese development of Small Modular Reactor Technology (ECC-SMART) collaborative project was established to encompass the design and pre-licensing requirements as well as a roadmap demonstrating the safe operation of the supercritical water small modular reactor (SCW-SMR).
One of the main challenges in the material and component aspect is the selection and qualification of a fuel cladding material that can withstand supercritical water conditions (beyond 374 °C and 22.1 MPa). The aim of the materials testing work package (WP2) in the ECC-SMART project is to achieve a deep understanding of the corrosion behavior of selected candidate cladding materials. Over 750 corrosion specimens were tested including those under nominal SCW-SMR operating conditions and also at simulated accident conditions. This article summarizes the findings from the study of corrosion behavior of non-irradiated and pre-irradiated candidate materials and from the study of the effect of chemistry and changes in the chemical properties of SCW.
{"title":"A state-of-the-art review of R&D for the supercritical water-cooled reactor technology. Part II materials & chemistry","authors":"K. Khumsa-Ang , M. Fulger , A. Sáez Maderuelo , A. Toivonen , M. Sipova , D. Marusakova , L. Zhang , J. Macák , R. Novotny","doi":"10.1016/j.nucengdes.2026.114789","DOIUrl":"10.1016/j.nucengdes.2026.114789","url":null,"abstract":"<div><div>This document presents a summary of the most relevant research and development (R&D) carried out to support the development of the only generation IV water-cooled reactor endorsed by the Generation IV International Forum (GIF).The coolant of the proposed reactor operates at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of high-grade process heat. Several collaborations have been established to support this technology under the GIF umbrella as well as through other international avenues; as a result, the development work is bolstered by a collective effort between numerous R&D institutions across Asia, Europe, and North America. The Joint European Canadian Chinese development of Small Modular Reactor Technology (ECC-SMART) collaborative project was established to encompass the design and pre-licensing requirements as well as a roadmap demonstrating the safe operation of the supercritical water small modular reactor (SCW-SMR).</div><div>One of the main challenges in the material and component aspect is the selection and qualification of a fuel cladding material that can withstand supercritical water conditions (beyond 374 °C and 22.1 MPa). The aim of the materials testing work package (WP2) in the ECC-SMART project is to achieve a deep understanding of the corrosion behavior of selected candidate cladding materials. Over 750 corrosion specimens were tested including those under nominal SCW-SMR operating conditions and also at simulated accident conditions. This article summarizes the findings from the study of corrosion behavior of non-irradiated and pre-irradiated candidate materials and from the study of the effect of chemistry and changes in the chemical properties of SCW.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114789"},"PeriodicalIF":2.1,"publicationDate":"2026-01-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078928","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 : 2026-01-24DOI: 10.1016/j.nucengdes.2026.114781
Zelong Zhao , Honghang Chi , Yuchen Xie , Yahui Wang , Yu Ma
Ultra-real-time simulation is crucial for ensuring the safe operation and control of nuclear power plants, as it enables rapid prediction and response to thermal-hydraulic behavior under accident conditions. This study proposes an ultra-real-time thermal-hydraulic modeling approach for the reactor primary circuit based on an intrusive reduced-order model (ROM). The governing equations of all components are discretized using a finite difference scheme, and variables for establishing ROMs are selected from these discretized equations to eliminate nonlinear terms. The transient solutions obtained from the initial 5% of time steps, calculated by the full-order model, served as snapshots, from which characteristic modes are extracted using the proper orthogonal decomposition. By projecting the discretized governing equations of each component onto the characteristic mode space, ultra-real-time thermal-hydraulic ROMs are constructed for each component. The integration of these ROMs for all components resulted in a comprehensive ultra-real-time model (URTM) of the primary circuit, capable of predicting system evolution. Simulation results demonstrated that the URTM achieves ultra-real-time performance while maintaining a maximum relative error of less than 0.15% for key thermal-hydraulic parameters.
{"title":"Ultra-real-time model reduction for nuclear reactor primary circuit calculation","authors":"Zelong Zhao , Honghang Chi , Yuchen Xie , Yahui Wang , Yu Ma","doi":"10.1016/j.nucengdes.2026.114781","DOIUrl":"10.1016/j.nucengdes.2026.114781","url":null,"abstract":"<div><div>Ultra-real-time simulation is crucial for ensuring the safe operation and control of nuclear power plants, as it enables rapid prediction and response to thermal-hydraulic behavior under accident conditions. This study proposes an ultra-real-time thermal-hydraulic modeling approach for the reactor primary circuit based on an intrusive reduced-order model (ROM). The governing equations of all components are discretized using a finite difference scheme, and variables for establishing ROMs are selected from these discretized equations to eliminate nonlinear terms. The transient solutions obtained from the initial 5% of time steps, calculated by the full-order model, served as snapshots, from which characteristic modes are extracted using the proper orthogonal decomposition. By projecting the discretized governing equations of each component onto the characteristic mode space, ultra-real-time thermal-hydraulic ROMs are constructed for each component. The integration of these ROMs for all components resulted in a comprehensive ultra-real-time model (URTM) of the primary circuit, capable of predicting system evolution. Simulation results demonstrated that the URTM achieves ultra-real-time performance while maintaining a maximum relative error of less than 0.15% for key thermal-hydraulic parameters.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114781"},"PeriodicalIF":2.1,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036028","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 : 2026-01-24DOI: 10.1016/j.nucengdes.2026.114759
Honglong Li , Xinxiang Long , Yunxin Zhang , Donghao He , Xiaojing Liu
FLASH, based on the fission response function (FRF) theory, is a high-fidelity and low-cost adaptive neutronics code. In this study, FLASH is validated in the AP1000 pressurized water reactor whole-core problem at hot zero power condition. It features a highly heterogeneous core arrangement with enrichment zoning, WABA and IFBA utilization. To accurately simulate the reactor core, an FRF database comprising 20 distinct assembly types was developed, and environmental factors were employed to account for local effects. Compared with Monte Carlo reference calculation, the FLASH AP1000-3D whole-core calculation yielded a error of 236 pcm and a RMS error in the pin-wise fission rate distribution of 0.99%. The entire core calculation was completed in less than 2 min.
{"title":"The validation of fission response function neutron transport code FLASH in AP1000 reactor core","authors":"Honglong Li , Xinxiang Long , Yunxin Zhang , Donghao He , Xiaojing Liu","doi":"10.1016/j.nucengdes.2026.114759","DOIUrl":"10.1016/j.nucengdes.2026.114759","url":null,"abstract":"<div><div>FLASH, based on the fission response function (FRF) theory, is a high-fidelity and low-cost adaptive neutronics code. In this study, FLASH is validated in the AP1000 pressurized water reactor whole-core problem at hot zero power condition. It features a highly heterogeneous core arrangement with enrichment zoning, WABA and IFBA utilization. To accurately simulate the reactor core, an FRF database comprising 20 distinct assembly types was developed, and environmental factors were employed to account for local effects. Compared with Monte Carlo reference calculation, the FLASH AP1000-3D whole-core calculation yielded a <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>e</mi><mi>f</mi><mi>f</mi></mrow></msub></math></span> error of 236 pcm and a RMS error in the pin-wise fission rate distribution of 0.99%. The entire core calculation was completed in less than 2 min.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114759"},"PeriodicalIF":2.1,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036029","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 : 2026-01-24DOI: 10.1016/j.nucengdes.2026.114794
Jehee Lee , Seong-Su Jeon , Ju-Yeop Park , Hyoung Kyu Cho
{"title":"Corrigendum to “Development of robustness assessment methodology for passive safety system against potential performance issue” [Nucl. Eng. Des. 449 (2026) 114755]","authors":"Jehee Lee , Seong-Su Jeon , Ju-Yeop Park , Hyoung Kyu Cho","doi":"10.1016/j.nucengdes.2026.114794","DOIUrl":"10.1016/j.nucengdes.2026.114794","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114794"},"PeriodicalIF":2.1,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189365","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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