Nuclear Engineering and Design最新文献

{"title":"Enhancing nuclear cogeneration efficiency using the low-grade waste heat recovery from nuclear hydrogen production system","authors":"Mehran Abolghasem ,&nbsp;Khashayar Sadeghi ,&nbsp;Seyed Hadi Ghazaie ,&nbsp;Ekaterina Sokolova ,&nbsp;Vitaly Sergeev ,&nbsp;Naypak Ksenia ,&nbsp;Wei Peng","doi":"10.1016/j.nucengdes.2025.114166","DOIUrl":"10.1016/j.nucengdes.2025.114166","url":null,"abstract":"<div><div>This study investigates the integration of modular high-temperature steam electrolysis (HTSE) into a nuclear power plant (NPP) to enhance cogeneration efficiency through low-grade waste heat recovery. Three integration scenarios are proposed, focusing on changing the discharge points within the NPP second cycle to minimize the power loss factor (PLF) and maximize overall system efficiency. Using Aspen HYSYS, detailed simulations were conducted to evaluate the thermodynamic performance of each scenario, while a PLF-based economic model developed to calculate the levelized cost of hydrogen (LCOHY) in each scenario. The results demonstrate that discharging low-grade steam after the last high-pressure preheater (Scenario 3) yields the highest cogeneration efficiency (38%) and the lowest LCOHY at 1.74 $/kg for large-scale systems. This scenario also achieves a 36.7% reduction in heat cost compared to the baseline configuration, which shows the economic and technical superiority of this scenario. The study reveals that large-scale HTSE systems outperform small-scale configurations, with lower PLF (36%) and higher scalability. By integrating waste heat recovery and optimizing steam return points, this work provides a novel framework for improving nuclear-hydrogen cogeneration, contributing to sustainable energy systems and the global transition to net-zero emissions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114166"},"PeriodicalIF":1.9,"publicationDate":"2025-05-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144084328","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":"Research on burnup solution method based on adaptive time step size","authors":"Cong Zhang ,&nbsp;Ling Chen ,&nbsp;Yongfa Zhang ,&nbsp;Song Li ,&nbsp;Binhang Zhang","doi":"10.1016/j.nucengdes.2025.114152","DOIUrl":"10.1016/j.nucengdes.2025.114152","url":null,"abstract":"<div><div>The time step size directly affects the accuracy and efficiency of burnup calculation. The division of the burnup step size depends on empirical knowledge, lacking a definitive theoretical foundation. In order to balance the influence of step size on calculation accuracy and efficiency, a burnup solution method based on adaptive time step size is developed in this work. In this research, the feasibility of the fast burnup solution method, which can solve the nucleon density at different burnup times using the same burnup matrix, based on the Mini-Max Polynomial Approximation (MMPA) is verified both theoretically and numerically. On this basis, a theoretical model for the division of burnup time step size is established based on the conversion relation between thermal power and neutron flux. Optimal step size is determined by iterative calculation to minimize the influence of the assumption of constant neutronics parameters on the calculation accuracy in the current time step and realize the adaptive discretization of the burnup step size during the whole life of the reactor. Furthermore, combined with the semi-predictor–corrector coupling strategy, an adaptive burnup step coupling strategy based on the MMPA method is proposed. Finally, the MOX fuel pin-cell and BWR assembly benchmarks are used for verification. The calculated results agree with the reference values, proving the correctness and effectiveness of the adaptive time step size method for solving burnup equations based on MMPA. This work provides a theoretical basis for the division of burnup time step size.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114152"},"PeriodicalIF":1.9,"publicationDate":"2025-05-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144070780","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":"Effect of non-spherical morphology on aerosol evolution in reactor containment using size-dependent dynamic shape factors","authors":"Jayant Krishan ,&nbsp;S. Anand ,&nbsp;A. Singh ,&nbsp;T. Thajudeen ,&nbsp;Jyoti Seth ,&nbsp;Y.S. Mayya","doi":"10.1016/j.nucengdes.2025.114136","DOIUrl":"10.1016/j.nucengdes.2025.114136","url":null,"abstract":"<div><div>The aerodynamic behavior of aerosol particles is significantly influenced by their size distribution and morphology. Conventional aerosol dynamics models often assume spherical, fully dense particles, which can lead to inaccuracies in predicting aerosol transport and deposition. This study highlights the necessity of incorporating dynamic shape factors to account for the non-sphericity of aerosol aggregates across different momentum transfer regimes. Using a geometric descriptors-based approach (GDA), we derive empirical relations for size-dependent dynamic shape factors, covering the entire Knudsen number range. These relations provide precise inputs for aerosol dynamics codes, enhancing the accuracy of nuclear accident simulations.</div><div>Application of these relationships to a pressurized heavy water reactor (PHWR) containment environment demonstrates that assuming spherical morphology underestimates airborne aerosol concentrations, particularly for aggregates with lower fractal dimensions. The study reveals that fractal aggregates experience increased drag forces, leading to significantly slower mass concentration reduction—approximately 25 times slower than that of spherical particles in a typical release scenario. By integrating these empirical expressions into aerosol modeling frameworks, this work improves predictions of aerosol transport, deposition, and airborne concentrations, thereby refining radiological safety assessments. The findings have direct implications for source term estimation, emergency response planning, and inhalation toxicology studies, reinforcing the critical role of dynamic shape factors in aerosol modeling.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114136"},"PeriodicalIF":1.9,"publicationDate":"2025-05-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144068828","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":"Basalt-boron fiber reinforced concrete: A sustainable solution for neutron radiation shielding","authors":"Maryna Holiuk ,&nbsp;Iryna Romanenko ,&nbsp;Hennadii Odynokin ,&nbsp;Anatolii Nosovskyi ,&nbsp;Vitalii Pastsuk ,&nbsp;Madis Kiisk ,&nbsp;Olexander Biland ,&nbsp;Anthony DiBenedetto ,&nbsp;Yurii Chuvashov ,&nbsp;Iryna Diduk ,&nbsp;Volodymyr Gulik","doi":"10.1016/j.nucengdes.2025.114150","DOIUrl":"10.1016/j.nucengdes.2025.114150","url":null,"abstract":"<div><div>The development of advanced composite materials is essential for addressing the growing demands of nuclear energy, particularly for radiation shielding. This study investigates the potential of basalt-boron fiber reinforced concrete as an innovative material to enhance the neutron shielding properties of concrete. Basalt-boron fiber reinforced concrete combines the mechanical benefits of basalt fiber with the neutron absorption capabilities of boron-10, offering a promising solution for nuclear reactor shielding applications. Experimental evaluations using a Pu-Be neutron source under two configurations – with and without cadmium nozzles – demonstrate that even low basalt-boron fiber dosages result in a neutron flux reduction of approximately 5% compared to plain concrete. Complementary Serpent Monte Carlo simulations validated these findings and further suggested that improved basalt-boron fiber production and increased fiber dosage can significantly enhance shielding performance.</div><div>The findings reveal that basalt-boron fiber reinforced concrete not only improves neutron shielding but also addresses critical ageing issues associated with biological shielding concrete in nuclear reactors. The uniform distribution of boron-10 within the fiber structure enables enhanced neutron absorption, particularly in the near-surface layers of biological shielding, where neutron-induced degradation is most severe. Furthermore, the fiber reinforcement mitigates microcracking, enhances fracture toughness, and prolongs the operational lifespan of reactor shielding structures. This feature is particularly crucial for nuclear power plants, where the integrity of biological shielding directly impacts safety and operational efficiency.</div><div>This study highlights the potential of basalt-boron fiber reinforced concrete as a cost-effective, durable, and scalable neutron shielding material for nuclear energy applications. The results establish a foundation for its broader application in nuclear reactor shielding and radiation protection systems, addressing both current and future challenges in the field.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114150"},"PeriodicalIF":1.9,"publicationDate":"2025-05-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144068829","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":"Neutronic and thermal performance analysis of advanced cladding materials for ACP-100 SMR","authors":"Md. Abidur Rahman Ishraq ,&nbsp;Afroza Shelley ,&nbsp;Md. Rashed Sardar","doi":"10.1016/j.nucengdes.2025.114142","DOIUrl":"10.1016/j.nucengdes.2025.114142","url":null,"abstract":"<div><div>This study analyzes the impact of various cladding materials: silicon carbide (SiC), Iron-Chromium-Aluminum Alloy (FeCrAl), Titanium-Molybdenum-Zirconium Alloy (TMZ), and Austenitic Stainless Steel (Alloy 33) and thicknesses (from 100 to 1000 µm) on the neutronic and thermal performance of the ACP-100 small modular reactor (SMR). Zircaloy is used as the reference material for comparison. Computational simulations using the Monte Carlo code SERPENT and nuclear data library ENDF/B-VII.1 indicate that SiC achieves the highest effective multiplication factor (keff) (1.3096), a 0.65 % increase over Zircaloy, while Alloy 33 exhibits the lowest keff (1.1724), a 9.90 % decrease at beginning of life (BOL) for thickness of 570 μm. As cladding thickness increases from 100 to 1000 µm, keff decreases by 4 % for SiC, 5.2 % for Zircaloy, and 18.2 % for Alloy 33 at BOL due to higher thermal neutron absorption. SiC sustains a cycle length of over 900 effective full power days (EFPDs) at 100 μm, achieving 912 EFPDs at 570 μm. In contrast, Alloy 33 shows poor neutron economy, with the cycle length dropping to 200 EFPDs at 1000 μm. SiC and TMZ demonstrate superior thermal conductivity (122.0 W/m·K and 110.3 W/m·K), reducing the fuel temperature by ∼10 K. While thicker cladding improves structural integrity, it compromises thermal efficiency. A thickness of 570 µm provides an optimal balance between performance and durability. SiC emerges as the most promising alternative cladding material for the ACP-100 SMR and further studies are recommended to assess its long-term behavior under reactor conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114142"},"PeriodicalIF":1.9,"publicationDate":"2025-05-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144068916","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 a separate effects MiniFuel irradiation experiment investigating microstructure evolution in high burnup UO2","authors":"Jacob P. Gorton,&nbsp;Annabelle G. Le Coq,&nbsp;Amber C. Telles,&nbsp;Kara M. Godsey,&nbsp;Jonathan R. Chappell,&nbsp;Christian M. Petrie,&nbsp;Nathan A. Capps","doi":"10.1016/j.nucengdes.2025.114149","DOIUrl":"10.1016/j.nucengdes.2025.114149","url":null,"abstract":"<div><div>The microstructural evolution of UO₂ fuel pellets during commercial operation in light water reactors (LWRs) is known to vary significantly across the pellet radius due to spatial variations in local temperature and burnup. The primary obstacle to extending LWR refueling cycles to 24-month intervals is the susceptibility of certain high burnup fuel microstructures to fuel fragmentation, relocation, and dispersal (FFRD) during a loss of coolant accident (LOCA). Although FFRD of the high burnup structure in the rim region of a pellet is well studied, the fine fragmentation that has been observed in a second region, near the midradius of the pellet (termed the “dark zone”) following mock LOCA testing of high burnup commercial fuel rods is less understood. This paper describes the design, analysis, and execution of a separate effects MiniFuel irradiation experiment that aims to identify the specific temperature and burnup regimes under which FFRD-susceptible dark zone microstructures form. The small disc specimens (3 mm diameter by ∼0.3 mm thick) enable more precise control of the relatively uniform temperature and burnup conditions. A total of 42 specimens were fabricated with typical LWR fuel densities (∼96%–98% of theoretical density) and grain sizes (∼12 μm) and are being irradiated over a range of temperatures (600°C–1000°C) and discharge burnups (50–72 MWd/kg-U) that bound the midradius region of high burnup LWR fuel. Fuel specimens with identical ²³⁵U enrichments were inserted in two irradiation locations in the High Flux Isotope Reactor and are currently undergoing irradiation to further evaluate the impact of rate effects (fission rate, time at temperature) on the microstructural evolution. The fuel fabrication and the thermal and neutronic simulations used for designing the experiment are detailed in this paper. A secondary objective of the experiment is to observe fission gas release (FGR) under the various irradiation conditions, and this work provides first-order predictions of FGR from all fuel specimens. The insights gained from these experiments will inform future high burnup core designs that could minimize the formation of susceptible microstructures and ultimately enable 24-month refueling cycles while minimizing the fraction of the fuel susceptible to FFRD.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114149"},"PeriodicalIF":1.9,"publicationDate":"2025-05-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144068830","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":"Study of thermal stratification in the upper plenum of the liquid metal fast reactor under mixed convection – An LES of the flow in the E-SCAPE facility","authors":"Ashish Saxena ,&nbsp;Matthew Falcone ,&nbsp;Shuisheng He","doi":"10.1016/j.nucengdes.2025.114112","DOIUrl":"10.1016/j.nucengdes.2025.114112","url":null,"abstract":"<div><div>A numerical analysis of thermal stratification in the upper plenum of the European SCAled Pool Experiment (E-SCAPE) facility was conducted using Large Eddy Simulation (LES) to enhance the understanding of thermal hydraulics phenomena in liquid metal fast reactors (LMFR) under mixed convection conditions. The geometric complexity in the above-core structure region was represented using a porous medium model. The results provide insights into the overall flow phenomena in the upper plenum region, the thermal instability in the above-core structure region, as well as the characteristics of rounded jets that emerge from the barrel walls. Under mixed convection conditions (low flow rate conditions), the strong buoyancy causes hot lead–bismuth eutectic (LBE) to accumulate at the top first, then flowing downwards, and then exiting the region through the upper set of barrel holes. Conversely, unmixed cold LBE spreads through the lower set of barrel holes. This results in a stratified temperature distribution, with lower temperatures at the bottom, slightly higher temperatures in the middle, and the highest temperatures at the top, demonstrating thermal stratification in the upper plenum region. This stratification occurs because the jets are weak in strength, resulting in poor mixing in the upper plenum region. Flow movements are confined to regions close to the jets, while areas away from the jets experience almost no movements or negligible movements, referred to as dead zones. It is useful to note that strong circulations are observed in one of our previous studies in the same facility under forced condition, which results in good mixing and no thermal stratification. In the above-core structure region, large-scale temperature fluctuations in the form of Kelvin-Helmholtz (KH) instabilities and mixing layers have been observed when the hot fluid backflows and interacts with the cold fluid.</div><div>The behaviour of the flows in the upper plenum region is dominated by the influence of different types of jets, including horizontally issued jets, jets angled upwards, and jets impinging on the components of the upper plenum. Under mixed convection conditions, the jets behave similarly to negatively inclined positively buoyant jet with minimal interactions between the top and middle jets and between the middle and bottom jets, occurring only in the vicinity of the jets. The latter stages of the jets indicate very little or almost negligible background flow movement. Additionally, high turbulence is observed in the shear layer of the jet orifice, which transitions into mixing layers after six jet diameters along the trajectory for the upper set of jets.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114112"},"PeriodicalIF":1.9,"publicationDate":"2025-05-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143947813","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":"Numerical investigation on characteristics of fully developed nucleate boiling (FDB) in petal-shape fuel assembly channel","authors":"Lipeng Du ,&nbsp;Tiezheng Wang ,&nbsp;Qi Cheng ,&nbsp;Xiang Chen ,&nbsp;Wenchao Zhang ,&nbsp;Jianchuang Sun ,&nbsp;Weihua Cai","doi":"10.1016/j.nucengdes.2025.114140","DOIUrl":"10.1016/j.nucengdes.2025.114140","url":null,"abstract":"<div><div>Fully developed nucleate boiling (FDB) distinguishes between single-phase flow and two-phase flow. In downstream of FDB, two-phase flow occurs. FDB is important for subcooled boiling. In this paper, a numerical method combining the Eulerian two-fluid model and the RPI wall boiling model was adopted to numerically investigate the heat transfer characteristics of coolant flow boiling in a 3 × 3 petal-shaped fuel assembly channel and studied on the characteristics of FDB. If the diameter of vapor bubble exceed the departing bubble diameter, the bubble can depart from the wall, FDB is determined. The results showed that the heating power had a minor effect on the characteristic of FDB. As the inlet subcooling increased, the temperature of mainstream on the cross section of FDB decreased and wall superheat increased. As the inlet mass flow rate increased, the mainstream temperature on the section of FDB increased and the wall superheat increased. With the pressure increase, the change trend of mainstream temperature and wall superheat was the same as mass flow rate increased. Studied the suitability of existing prediction correlations, and the result showed that the error of Unal’s expression prediction of FDB was the minimum. To decrease the error, a factor was introduced to modify the prediction expression. Comparing with simulation result, the error of modified expression was less than 15 %. The wall temperature at the inner concave arc was much higher than that at the outer convex arc, and the intensity of the coolant subcooled boiling was sharply at the inner concave arc.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114140"},"PeriodicalIF":1.9,"publicationDate":"2025-05-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143947812","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":"Insights on the steady-state performance of single-phase natural circulation loops","authors":"Swati Gangwar ,&nbsp;P.K. Vijayan ,&nbsp;Goutam Dutta","doi":"10.1016/j.nucengdes.2025.114128","DOIUrl":"10.1016/j.nucengdes.2025.114128","url":null,"abstract":"<div><div>The estimation of the induced flow rate is required to determine the heat transport capability of natural circulation loops. A generalized dimensionless equation for the steady-state flow for single-phase natural circulation loops is reported in the literature. This generalized equation applies to uniform and non-uniform diameter rectangular loops with a uniform heat flux boundary condition for the heater and a specified overall heat transfer coefficient and secondary side coolant temperature for the cooler. The applicability of this generalized equation is examined to other heater and cooler boundary conditions like specified power density, isothermal wall, heat transfer coefficient, and coolant temperature. Further, the applicability of the generalized equation is examined to loops of other shapes, such as toroidal, figure of eight loop, pentagon, and thermosyphon heat transport devices (THTDs). Natural circulation is possible with gravitational and centrifugal force fields. Hence, the applicability of the generalized equation to rotating natural circulation loops operating in the centrifugal force field is also examined. Besides, the applicability of the generalized equation to open natural circulation loops, interconnected circuits, and coupled natural circulation loops (CNCLs) is examined, along with its applicability to other fluids. In all cases, the available experimental data are used to test the validity of the dimensionless flow equation. Finally, the generalized equation is tested with reported CFD data. In all cases, a reasonable comparison is obtained. In addition, the reasons for the deviation between the data and the generalized equation are also presented.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114128"},"PeriodicalIF":1.9,"publicationDate":"2025-05-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143948573","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":"Integrating spatial prediction and Bayesian optimization for dynamic event tree optimization branch search","authors":"Haoyin Chen,&nbsp;He Wang,&nbsp;Longcong Wang,&nbsp;Qiang Zhao","doi":"10.1016/j.nucengdes.2025.114143","DOIUrl":"10.1016/j.nucengdes.2025.114143","url":null,"abstract":"<div><div>The dynamic event tree (DET) method has been extensively utilized in safety analyses of nuclear power plant accidents. However, due to the complex nature of accident response processes, ensuring a comprehensive simulation of accident outcomes and calculating precise failure probability values requires an enormous number of discrete branches generated by DET, which incurs a prohibitive computational cost. This paper proposes a DET branch search method that integrates spatial prediction and Bayesian optimization. This method uses a limited number of DET branch calculation results as prior data and employs Bayesian optimization to automatically search for the optimal branch step size. By integrating Dynamic Time Warping (DTW) classification with data stability testing, the status of unknown spaces is preliminarily determined, gradually reducing the scope of spaces with potential failure risks. Through the selection of critical yet minimal DET branches for a comprehensive simulation of accident outcomes, computational efficiency is enhanced, and accurate failure probability results are achieved. Using the operation time of the turbine-driven auxiliary feedwater system and the main pump shaft seal leakage time in a station blackout (SBO) accident of a CPR1000 nuclear power plant as an example, a DET model is constructed. Ultimately, the computational time and failure probability of the DET branching optimization search method are compared with those of the equally spaced variable values method. The relative error of the failure probability is within 1.96%, while computational efficiency is improved by a factor of 4.82. This approach not only ensures computational accuracy but also enhances the efficiency of DET calculations, offering substantial methodological support for the practical application of DET.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"440 ","pages":"Article 114143"},"PeriodicalIF":1.9,"publicationDate":"2025-05-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143947810","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}