Nuclear Engineering and Design最新文献

{"title":"Gas-liquid flow characteristics through a Micron scale orifice of failed fuel pin in Lead-bismuth cooled reactors","authors":"Yuchen Li,&nbsp;Yanmin Zhou,&nbsp;Haifeng Gu,&nbsp;Yichen Zhang,&nbsp;Shuolei Fan","doi":"10.1016/j.nucengdes.2026.114772","DOIUrl":"10.1016/j.nucengdes.2026.114772","url":null,"abstract":"<div><div>Fuel cladding in lead‑bismuth eutectic (LBE) cooled reactors may develop micron-scale failures during long-term high-temperature operation. The flow regime of gas submerged jet at the defect orifice directly influences the scrubbing behavior of these fission products. This study investigated a micron-scale failures in LBE system by conducting visualization experiments using inert gas and deionized water as working fluids. The research first systematically compared the criteria for bubbling-jet regime transition between micron-scale and millimeter-scale orifices. The results revealed that the traditional criteria based on the liquid-phase Weber number or Mach number, which are applicable for millimeter-scale orifices failed at micron scale. The phenomenon was primarily attributed to the lower gas momentum from micron-scale orifice, which allowed for continued flow regime evolution even after critical flow was reached. Consequently, a new criterion defined as the product of a density correction factor and the liquid Weber number was proposed to accurately predict the flow regime transition. Furthermore, an empirical correlation for predicting the Sauter mean diameter of the gas bubbles was established based on the SPARC90 bubble size prediction model, with a prediction error within the range of −15% to +15%. Finally, the results were extrapolated to a prototypical lead‑bismuth environment through scaling analysis, which verified a good predictive capability for bubble size generated from micron-scale orifices of this correlation. The research provided theoretical support for two-phase flow regime identification and bubble behavior prediction in the safety analysis of lead‑bismuth reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114772"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036026","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":"Ultra-real-time model reduction for nuclear reactor primary circuit calculation","authors":"Zelong Zhao ,&nbsp;Honghang Chi ,&nbsp;Yuchen Xie ,&nbsp;Yahui Wang ,&nbsp;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":"2026-04-01","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}

{"title":"Design of 316L-based neutron shielding materials and preparation and characterization of Gd/316L","authors":"Chunguang Qiao ,&nbsp;Zhonghua Wang ,&nbsp;Xinpeng Wei ,&nbsp;Dehui Wu ,&nbsp;Chao Jiang","doi":"10.1016/j.nucengdes.2026.114808","DOIUrl":"10.1016/j.nucengdes.2026.114808","url":null,"abstract":"<div><div>The corrosion resistance and structural mechanical properties of neutron shielding materials such as B₄C/Al and boron‑aluminum alloy are insufficient. The 316L stainless steel commonly used in the nuclear industry has excellent structural mechanical properties and corrosion resistance, but its neutron shielding performance is poor. The above-mentioned materials are unable to meet the multi-dimensional functional requirements of the core shielding layer of fast reactors, the structural components within the reactor, and the storage of spent fuel, which include high load-bearing capacity, efficient neutron shielding, and excellent corrosion resistance. Therefore, 316L-based neutron shielding materials are designed and prepared using 316L stainless steel as the substrate and B₄C, Gd, Sm₂O₃, and Eu₂O₃ as neutron shielding enhancers. Firstly, the relationships between the areal densities of B₄C, Gd, Sm₂O₃, and Eu₂O₃, and the neutron shielding rate and secondary γ-ray production rate of the corresponding shielding materials are established. Secondly, using the established relational equation, the required contents of B₄C, Gd, Sm₂O₃, and Eu₂O₃ in 316L stainless steel are calculated and compared using the neutron shielding rate of 30% B₄C/Al composite as the design basis. Finally, Gd/316L neutron shielding materials are prepared by directed energy deposition additive manufacturing (DED-AM) process, and their micro-morphology and mechanical properties are analyzed. The results show that Gd is more suitable as a neutron shielding enhancer for 316L stainless steel. The 1.9% Gd/316L exhibits good mechanical properties, while a further increase in Gd content degrades the mechanical performance of the material</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114808"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190736","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":"Development of robustness assessment methodology for passive safety system against potential performance issue","authors":"Jeehee Lee ,&nbsp;Seong-Su Jeon ,&nbsp;Ju-Yeop Park ,&nbsp;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":"2026-04-01","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}

{"title":"Multidimensional optimization of the high-diodicity diaphragm hydrodiode for passive safety systems of nuclear power plants","authors":"Victor Shcherba ,&nbsp;Anatoliy Khait ,&nbsp;Sergey Kaigorodov ,&nbsp;Ksenia Sokirko ,&nbsp;Evgeniy Pavlyuchenko","doi":"10.1016/j.nucengdes.2026.114803","DOIUrl":"10.1016/j.nucengdes.2026.114803","url":null,"abstract":"<div><div>A novel high-efficiency diaphragm hydrodiode (i.e., fluidic diode) for NPP safety circuits is proposed. To achieve maximum diodicity, a multi-parameter optimization of its geometry is performed using a machine-learning-aided surrogate model. Training the surrogate model is performed using the quasi-random sampling, while the exact diodicity values were provided by the CFD simulations based on the Reynolds-Averaged Navier-Stokes equations closed with the <span><math><mi>k</mi><mo>−</mo><mi>ω</mi></math></span> turbulence model. Iterative complementation of the sampling is employed to further increase the surrogate model accuracy. Genetic and Trust-Region optimization algorithms are executed on top of the surrogate model to arrive at the optimal hydrodiode configuration. The maximum diodicity value reported by both CFD and the surrogate model is <span><math><msub><mi>D</mi><mi>CFD</mi></msub><mo>≈</mo><mn>2.74</mn></math></span>, while the experimentally confirmed diodicity of the optimal diode configuration is found to be <span><math><msub><mi>D</mi><mi>exp</mi></msub><mo>=</mo><mn>2.59</mn></math></span>. Such a high diodicity value for the diaphragm hydrodiode is reported for the first time, thus constituting an achievement in the field. The proposed design and optimization methodology open up possibilities for constructing compact and reliable passive components for safety systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114803"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081253","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":"Initial study on the effects of TRISO-matrix interactions on TRISO particle performance in ceramic matrix dispersion microencapsulated fuel pellet","authors":"Junqiang Zheng ,&nbsp;Jingyu Nie ,&nbsp;Ya'nan He ,&nbsp;Yingwei Wu ,&nbsp;Jing Zhang ,&nbsp;Yuanming Li ,&nbsp;Shichao Liu ,&nbsp;G.H. Su","doi":"10.1016/j.nucengdes.2026.114811","DOIUrl":"10.1016/j.nucengdes.2026.114811","url":null,"abstract":"<div><div>This study presents a comprehensive multi-physics coupled analysis of TRISO-based Ceramic matrix Dispersion Microencapsulated (CDM) fuels, focusing on the debonding behavior at the TRISO-matrix interface under irradiation and thermal conditions, as well as the influence of TRISO-matrix interactions on TRISO particle performance within fuel pellets. A two-dimensional axisymmetric representative unit cell model was developed to investigate the effects of interfacial debonding characteristics under varying bonding parameters. Based on these findings, a three-dimensional pellet-scale multi-physics coupled model incorporating the Cohesive Zone Model was established to simulate interfacial debonding between TRISO particles and the matrix and its impact on structural integrity and failure probability. The model integrates irradiation-induced deformation, thermomechanical behavior, fission gas release, and interfacial damage evolution, enabling detailed evaluation of stress distribution and failure probability in both TRISO particles and SiC matrices.</div><div>Key results reveal that early-stage debonding at the TRISO-matrix interface is primarily driven by thermal expansion mismatch, with irradiation-induced deformation of the PyC layer contributing minimally. Smaller initial damage displacements accelerate interface failure, while larger failure displacements delay debonding but increase coating layer stresses, particularly in the SiC layer. At low temperatures, internal gas pressure within TRISO particles remains insufficient to induce critical tensile stress in the SiC layer, whereas PyC layer irradiation-induced deformation dominates the failure probability of the SiC coating. Additionally, spatial particle distribution and the surrounding matrix state significantly influence SiC coating failure, with localized high tensile stresses in narrow matrix regions posing risks for microcrack initiation.</div><div>Comparative analysis between pellet-scale CDM models and single TRISO particle models highlights the amplification of PyC layer hoop stresses due to particle-matrix interactions. The maximum failure probability of SiC coatings in pellet-scale models is approximately one order of magnitude higher than in single-particle simulations, underscoring the critical role of matrix confinement and interparticle interactions. These findings emphasize the necessity of considering high-packing-fraction configurations and interfacial bonding parameters in TRISO-based fuel performance assessments. Future work will focus on quantifying the effects of particle-matrix interactions on TRISO particle performance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114811"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190259","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":"Near-wall void distribution characterization in pebble bed reactor using gamma-ray CT and DEM simulation","authors":"Ahmed Jasim ,&nbsp;Mauricio Maestri ,&nbsp;Abdullah Al Zubaidi ,&nbsp;Omar Farid ,&nbsp;Muthanna Al-Dahhan","doi":"10.1016/j.nucengdes.2026.114813","DOIUrl":"10.1016/j.nucengdes.2026.114813","url":null,"abstract":"<div><div>Accurate characterization of void-fraction distributions in pebble-bed reactors (PBRs) is essential for predicting flow, heat transfer, and neutronic behavior. High-fidelity experimental benchmark data for validating such predictions remain scarce, largely due to the challenges of non-invasive measurements. In this study, gamma-ray computed tomography (CT) was employed to measure radial and cross-sectional porosity in a laboratory-scale pebble bed containing <span><math><mrow><mn>6</mn><mspace></mspace><mi>cm</mi></mrow></math></span> graphite pebbles. A Discrete Element Method (DEM) simulation was implemented and validated against these measurements, then applied to the full-scale geometry of the Xe-100 high-temperature gas-cooled pebble-bed reactor. Analyses included radial and axial void-fraction profiles in the cylindrical section and conical base, with particular attention to near-wall oscillations at multiple axial levels. Both axially averaged profiles, integrating over extended bed sections, and locally resolved profiles, capturing fine-scale oscillations, were evaluated. Additional analyses examined cross-sectional void distributions and the effect of pebble recirculation. The DEM results reproduced the expected near-wall oscillatory layering with a characteristic wavelength of ∼1 <span><math><mrow><mi>dp</mi></mrow></math></span> and bulk void fractions near <span><math><mrow><mn>0.40</mn></mrow></math></span> and further showed that oscillatory patterns persist into the conical region, where the first trough shifts outward, and a broader near-wall gap develops. Recirculation studies, corresponding to <span><math><mrow><mn>5</mn><mo>,</mo><mn>10</mn><mo>,</mo></mrow></math></span> and <span><math><mrow><mn>15</mn></mrow></math></span> complete bed inventory cycles, showed that structural rearrangements occur mainly during the initial passes, after which the bed attains a quasi-steady configuration. Recirculation intensified near-wall oscillations, particularly in the lower regions, but had negligible impact on bulk porosity in the cylindrical section. In the cone, however, the void fraction was elevated during dynamic operation due to pebble drainage and upward void propagation. The findings support improved neutronic and thermal-hydraulic modeling and contribute to the design and safety assessment of next-generation pebble-bed systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114813"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190258","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":"Analytic hierarchy process-based integrated assessment of IAEA'S 19 infrastructure criteria for small modular reactor technology assessment","authors":"Alif Imran Mohd Shuhaimi,&nbsp;Mohd Syukri Yahya,&nbsp;Shamsul Amri Sulaiman,&nbsp;Eli Syafiqah Aziman","doi":"10.1016/j.nucengdes.2026.114797","DOIUrl":"10.1016/j.nucengdes.2026.114797","url":null,"abstract":"<div><div>Achieving Net Zero Emissions (NZE) by 2050 necessitates a strategic combination of renewable energy sources and dependable, clean baseload power. Nuclear energy is increasingly recognized as a critical element of this transition, with recent global policy endorsements calling for a tripling of nuclear capacity, underscoring its growing importance. Substantial expansion is anticipated not only in established nuclear markets but also in newcomer countries. However, initiating nuclear power programs poses significant challenges for newcomers, particularly due to massive capital investment, long-term infrastructure commitments, and the complexity of managing associated development risks. Reactor Technology Assessment (RTA) is therefore an essential process to ensure nuclear projects are delivered on time, within budget, safely, and to specifications. The International Atomic Energy Agency (IAEA) supports this process through its comprehensive framework of 19 infrastructure issues, addressing key dimensions such as national position, technical, and economic factors. This study presents a systematic, multi-criteria decision-making framework for conducting RTA, focusing on Small Modular Reactors (SMRs) as viable options for deployment in newcomer countries. The framework integrates the IAEA's infrastructure guidance with a PESTEL-based evaluation model. A case study is conducted for Malaysia to identify suitable SMRs for a Nuclear Hydrogen Hybrid System (NHHS). The assessment framework is validated using the Analytic Hierarchy Process (AHP), ensuring robustness and consistency in decision-making. The findings provide insights into aligning SMR technology selections with national requirements and risk profiles, offering a replicable methodology to support evidence-based decision-making in emerging nuclear energy programs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114797"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190256","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":"Advanced acoustically tensioned metastable fluid detectors for 10-15× over He-3 detector efficiency, real-time monitoring of moving SNMS, and spectroscopic identification of fission vs. alpha-n neutron sources","authors":"S. Ozerov ,&nbsp;N. Boyle ,&nbsp;D. DiPrete ,&nbsp;R.P. Taleyarkhan","doi":"10.1016/j.nucengdes.2026.114809","DOIUrl":"10.1016/j.nucengdes.2026.114809","url":null,"abstract":"<div><div>This paper addresses a need for spectroscopic monitoring for the illicit movement of stationary/moving special nuclear materials (SNMs) via high-efficiency real-time detection from tell-tale neutron radiation signatures, especially while in motion, and with 100% background gamma-beta blindness. Studies were conducted in a direct one-on-one comparison vs current state-of-the-art technology, which conventionally employs He-3 detectors, which require bulky shielding for neutron down-scattering. The advanced Economical Acoustically Tensioned Metastable Fluid Detector [<em>E</em>-ATMFD(Ver.1)] developed at Purdue University promises to offer 10× improved efficiency of SNM detection of equivalent form factor He-3 detectors. This paper details the design and performance of the advanced <em>E</em>-ATMFD(Ver.1) sensor system using borated sensing fluid, which demonstrated absolute efficiencies x10–20 greater than a state-of-the-art moderated Ludlum-42-49B (He-3) detector of equivalent size, when compared side-by-side for detecting and tracking a shielded and unshielded moving Plutonium-Beryllium (SNM surrogate of similar intensity) neutron source. An additional factor of 2× improvement was obtained by tailoring the internal and external static pressure fields, leading to enhanced dynamic tensioned (negative) pressure (Pneg) metastability states within the <em>E</em>-ATMFD(Ver.1). By sweeping the Pneg states, the E-ATMFD (Ver.1) was also shown capable of spectroscopically distinguishing a Cf-252 fission multiplicity-bearing neutron source from a non-multiplicity Am<img>Be (α,n) random neutron source.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114809"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190252","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":"Investigation on diversity of trip parameters in reactor protection system for SALUS under various anticipated operational occurrences","authors":"Ji-Woong Han,&nbsp;Sun Rock Choi,&nbsp;In Sub Jun,&nbsp;Seungjoon Baik,&nbsp;Huee-Youl Ye,&nbsp;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":"2026-04-01","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}