Nuclear Engineering and Design最新文献

{"title":"A multi-level sensitivity analysis for the model parameters of the main steam system in nuclear power plants","authors":"Chenke Ding ,&nbsp;Xiaoyu Luo ,&nbsp;Sheng Zheng ,&nbsp;Dazhi Zhang ,&nbsp;Xian Zhang ,&nbsp;Shanglong Huang ,&nbsp;Yanda Zhu ,&nbsp;Keming Ren ,&nbsp;Junjie He","doi":"10.1016/j.nucengdes.2025.114420","DOIUrl":"10.1016/j.nucengdes.2025.114420","url":null,"abstract":"<div><div>The main steam system of a nuclear power plant is a core component of its thermal system, and its operation is typically monitored using simulation models to ensure both efficiency and safety. However, the accuracy of the system model is influenced by the uncertainty of multiple parameters. In this context, sensitivity analysis is essential, as it identifies the most key model parameters, thereby reducing the parameter space and enhancing the efficiency and effectiveness of model calibration. This paper presents a multi-level sensitivity analysis framework that combines the Morris method and the Generalized Likelihood Uncertainty Estimation (GLUE) method. The Morris method is employed as an efficient preliminary screening technique to identify parameters that potentially exert significant influence on model outputs, thereby effectively reducing the dimensionality of the parameter space. Subsequently, the GLUE method is applied to assess the model’s goodness-of-fit to benchmark data using the Nash–Sutcliffe efficiency coefficient, and the posterior distribution of model parameters is obtained to quantify the importance of the parameters. The results show that eight model parameters significantly affect the model output, providing theoretical guidance for model optimization and parameter adjustment of the nuclear power plant’s main steam system. The proposed framework reduces computational time from 26.89 h to 10.73 h, improving efficiency by 60.1% while maintaining high accuracy in key model parameter identification compared to traditional approaches.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114420"},"PeriodicalIF":2.1,"publicationDate":"2025-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144931987","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 computationally effective models for simulation of steam injection effects on the pool stratification and mixing","authors":"Xicheng Wang ,&nbsp;Pavel Kudinov ,&nbsp;Dmitry Grishchenko ,&nbsp;Ralf Kapulla ,&nbsp;Sidharth Paranjape ,&nbsp;Domenico Paladino ,&nbsp;Simon Suter ,&nbsp;Markku Puustinen ,&nbsp;Antti Räsänen ,&nbsp;Lauri Pyy ,&nbsp;Eetu Kotro","doi":"10.1016/j.nucengdes.2025.114406","DOIUrl":"10.1016/j.nucengdes.2025.114406","url":null,"abstract":"<div><div>Steam injection through blowdown pipes and spargers into a large water pool, also known as Pressure Suppression Pool (PSP), is employed in Boiling Water Reactors (BWRs) to prevent containment overpressure. Thermal stratification in the pool results in an increased pool surface temperature compared to a mixed pool condition, leading to higher containment pressure. Therefore, adequately validated predictive capabilities for modeling of the pool behavior are essential for the safety analysis of containment performance. The thermal behavior of the pool (e.g. thermal stratification or mixing transient) depends on the interplay between the heat and momentum sources induced by direct contact condensation of steam. Computational efficiency of the models for the simulation of the long transients in the large-scale pools is critical, especially for the quantification of uncertainties. The Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been developed to represent the impact of steam injection on the pool while avoiding the detailed simulation of steam-water interface dynamics, which is a computational challenge in itself. These models are compatible with any Computational Fluid Dynamics (CFD) code using a single-phase solver. In this work we further develop the EHS/EMS models using (i)<!--> <!-->new EMS model correlation based on the latest results from Separate Effect Test (SEF-POOL) facility; and (ii)<!--> <!-->Condensation-Induced Turbulence (CIT) model calibrated against integral pool experiments conducted at PANDA, PPOOLEX, SJTU, and HEU facilities under a wide range of steam injection conditions. The good agreement of the global pool behavior and local flow characteristics demonstrates that the proposed models can provide an adequate prediction of the relevant phenomena.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114406"},"PeriodicalIF":2.1,"publicationDate":"2025-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144931988","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":"CATHARE calculation of the tests conducted on the ELSMOR passive heat removal system","authors":"B. Grosjean ,&nbsp;R. Ferri ,&nbsp;C. Lombardo","doi":"10.1016/j.nucengdes.2025.114437","DOIUrl":"10.1016/j.nucengdes.2025.114437","url":null,"abstract":"<div><div>As part of the ELSMOR (Toward European Licensing of Small Modular Reactors) project, an experimental facility has been set up at SIET (Piacenza, Italy) to test a passive heat removal system. Such a passive system operates in natural circulation, with three different circuits: a primary circuit (PC, the heat source), a secondary circuit (SC, self-pressurised) and a tertiary circuit (the heat sink, represented by a pool). The primary and the secondary circuits are thermally coupled to a plate-type Compact Steam Generator (CSG), while the secondary and the tertiary circuits are coupled to an in-pool condenser. An experimental campaign has been carried out to investigate the effect of different parameters on the passive system behavior, through different types of tests (e.g. secondary side filling ratio (FR) or non-condensable gas (NC) concentration, primary system temperature, pool level, etc.). These experimental tests are modelled with the CATHARE 3 code (French system calculation code) and the calculation results are compared with the experimental data. The CATHARE 3 code predicts good tendencies for the tests on the main parameters of the facility (exchanged power, secondary circuit pressure, condenser outlet temperature). For the majority of the tests, the discrepancy between experimental and calculation results for the exchanged power in the CSG is below 10 %: for high FR in the SC, the CATHARE 3 code predicts the exchanged power well, while for low FR the power is overestimated. Sensitivity calculations showed that the condenser has the main influence on the facility behavior (the CSG has a limited influence); in particular, the correlations in the secondary side of the condenser have a significant influence on the exchanged power, while the correlations in the tertiary side have a small influence; thus, the tertiary modelling has a small influence on the calculation results. For all tests, the CATHARE 3 code underestimates the SACO outlet temperature and overestimates the pressure in the SC. Nonetheless, the presence of experimental uncertainties, particularly related to uncharacterized head losses and two-phase flow conditions, prevents drawing fully conclusive statements about the model accuracy.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114437"},"PeriodicalIF":2.1,"publicationDate":"2025-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144925710","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":"Multiphysics coupling analysis of heat pipe reactor considering irradiation effects based on RMC and FEniCSx","authors":"Dacai Zhang ,&nbsp;Ningkun Zhang ,&nbsp;Zeguang Li ,&nbsp;Yushuo Ren ,&nbsp;Zilin Su ,&nbsp;Ganglin Yu ,&nbsp;Guanghui Zhong ,&nbsp;Haochun Ding","doi":"10.1016/j.nucengdes.2025.114414","DOIUrl":"10.1016/j.nucengdes.2025.114414","url":null,"abstract":"<div><div>In this study, a three-dimensional neutronics-thermal-mechanical multiphysics coupling process was developed based on the reactor Monte Carlo software RMC and the finite element software FEniCSx. The computational model accounts for the irradiation effects on material properties and incorporates the irradiation swelling model to analyze the KRUSTY heat pipe reactor. The results reveal that the negative reactivity induced by thermal expansion accounts for 90% of the total reactivity feedback in the zero-burnup condition. After 15 years burnup, the irradiation effects lead to increases of 11.19% and 10.63% in maximum displacement and stress, while introducing a negative reactivity feedback of 71.69 pcm. With 45 years of operation, the negative reactivity due to irradiation reaches 218 pcm accounting for 26. 86% of the total negative reactivity feedback, the displacement changes caused by the irradiation swelling account 35% of the thermal expansion. These results demonstrate that the irradiation swelling has a significant impact under high burn-up conditions, providing useful insights for the future design of the heat pipe reactor.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114414"},"PeriodicalIF":2.1,"publicationDate":"2025-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144916553","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":"Assessment of TRACE for predicting startup transients in natural circulation boiling water reactors","authors":"Yang Zhao,&nbsp;Yunlin Xu,&nbsp;Mamoru Ishii","doi":"10.1016/j.nucengdes.2025.114430","DOIUrl":"10.1016/j.nucengdes.2025.114430","url":null,"abstract":"<div><div>Natural circulation is a key passive safety feature in several light water small modular reactor (SMR) designs, but flow instabilities during natural circulation startup transients pose critical challenges for safety analysis. To ensure accurate prediction of key thermal–hydraulic phenomena in startup transients, system-level analysis codes must be rigorously assessed. Among these, TRACE is recognized as the flagship best-estimate reactor system code developed by the U.S. Nuclear Regulatory Commission (NRC). This study evaluates TRACE (V5p8) for predicting two-phase flow behavior in natural circulation using startup transient data from the Purdue University Multidimensional Integral Test Assembly (PUMA). Four startup tests (EQU, EQU-NU, SUP, SUP-NU) were analyzed, simulating natural circulation boiling water reactor (BWR) behavior under different pressure conditions, both with and without neutronic feedback. A systematic refinement study was conducted to determine the optimal TRACE simulation settings, including nodalization, timesteps, and spatial discretization methods. Results indicate that TRACE successfully handles input power oscillations without significant numerical instabilities and generally provides reasonable predictions. While TRACE accurately predicts system pressure and liquid temperature, it significantly underestimates core void fractions, failing to capture geysering phenomena due to underprediction of energy allocated for void generation in subcooled boiling flow. The flashing phenomenon in the chimney is well captured, but chimney void fractions are generally overestimated. Additionally, TRACE overestimates downcomer flow rates in most cases, potentially due to the overestimation of chimney void fraction, which influences natural circulation gravity head. Future work will focus on further validation of TRACE models using an extended natural circulation database from separate effect tests (SETs) and integral effect tests (IETs) to enhance modeling accuracy for complex startup transients in natural circulation systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114430"},"PeriodicalIF":2.1,"publicationDate":"2025-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144916554","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":"3-D numerical studies of single control rod withdrawal transients in an LBE cooled critical reactor","authors":"Xue-Nong Chen ,&nbsp;Yoshiharu Tobita ,&nbsp;Andrei Rineiski ,&nbsp;Barbara Kędzierska ,&nbsp;Guy Scheveneels ,&nbsp;Matteo Zanetti ,&nbsp;Bogdan Yamaji","doi":"10.1016/j.nucengdes.2025.114419","DOIUrl":"10.1016/j.nucengdes.2025.114419","url":null,"abstract":"<div><div>The presented studies are carried out within the EU project ANSELMUS. A recent design version of MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications), which is a Lead-Bismuth-Eutectic (LBE) cooled reactor developed at SCK-CEN (Belgian Nuclear Research Centre), is investigated. The SIMMER-IV code is employed for 3-D simulations of single control rod withdrawal (CRWD) transients at the critical operation mode. A new CRWD model for the SIMMER-IV code is developed, so that the CRWD can be simulated for any constant withdrawal speed and from any initial position. The basic case is the complete withdrawal of a control rod (CR) filled with B4C absorber with the natural boron, where the reactivity worth is about 0.9 $, from a fully inserted position within 3 s. Cases with and without scram after 3 s are considered. The spatial kinetics effects on the power distribution are evaluated by comparing relative variations in time of local power densities and of the total one. The dynamic reactivity values during CRWD have been confirmed to be close to those obtained by static calculations. The transient with the scram at 3 s results in nothing severe, but that without the scram leads to local fuel melting. A further example, where the CR is filled with an enriched by B-10 absorber, its reactivity worth being 1.7 $, is calculated and shown as well. The withdrawal leads to a severe accident with fuel pin degradation, but without prompt supercritical power excursion. The numerical scenarios are presented and investigated by means of parametric studies.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114419"},"PeriodicalIF":2.1,"publicationDate":"2025-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144920174","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 analysis of counter-current flow limitation in a horizontal hot-leg with elbow using RELAP5 and VOF–RANS models","authors":"Johan Sarache Piña ,&nbsp;Santiago Corzo ,&nbsp;Dario Godino ,&nbsp;Damian Ramajo","doi":"10.1016/j.nucengdes.2025.114416","DOIUrl":"10.1016/j.nucengdes.2025.114416","url":null,"abstract":"<div><div>During the <em>reflux-cooling</em> phase that follows certain small-break loss-of-coolant accidents (SBLOCAs) in pressurized water reactors (PWRs), the condensate formed in the steam generators must descend through the <em>hot-leg</em> to re-flood the core. This downward liquid stream can be throttled by the counter-current steam rising from the vessel, a phenomenon known as <strong>counter-current flow limitation</strong> (CCFL). A reliable CCFL prediction is therefore pivotal for estimating the passive cooling capability of the primary circuit and, in turn, for judging the safety margin in SBLOCA scenarios. CCFL is quantified here for the <em>COLLIDER</em> facility (190 mm ID) using two complementary strategies: (i) a three-dimensional VOF–RANS model with variable-density (<span><math><mi>ρ</mi></math></span>-var) turbulence formulation in <span>OpenFOAM</span> <!--> <!-->v2206, and (ii) the one-dimensional system code <span>RELAP5</span>-Mod3 with a linear Wallis-type flooding correlation. Four operating regimes are examined in the range <span><math><mrow><msubsup><mrow><mi>J</mi></mrow><mrow><mi>f</mi></mrow><mrow><mo>∗</mo><mn>0</mn><mo>.</mo><mn>5</mn></mrow></msubsup><mo>=</mo><mn>0</mn><mo>.</mo><mn>10</mn><mtext>–</mtext><mn>0</mn><mo>.</mo><mn>30</mn></mrow></math></span>. Results show that the VOF–RANS model reproduces the so-called <em>elevated CCFL</em> — i.e. a controlled overshoot of the Wallis line before full blockage — with errors below 10 % in both pressure drop and blockage onset. In contrast, <span>RELAP5</span> anticipates blockage by up to 25 %, confirming its conservative bias. Parametric studies reveal that the <span><math><mi>ρ</mi></math></span>-var formulation lowers the excess interfacial drag by roughly 40 % relative to the incompressible variant, and that mesh refinements finer than 10 mm produce marginal changes in global outcomes. Two practical guidelines emerge: (a) the 1-D approach is adequate for <span><math><mrow><msubsup><mrow><mi>J</mi></mrow><mrow><mi>f</mi></mrow><mrow><mo>∗</mo><mn>0</mn><mo>.</mo><mn>5</mn></mrow></msubsup><mo>&lt;</mo><mn>0</mn><mo>.</mo><mn>15</mn></mrow></math></span>; (b) for <span><math><mrow><msubsup><mrow><mi>J</mi></mrow><mrow><mi>f</mi></mrow><mrow><mo>∗</mo><mn>0</mn><mo>.</mo><mn>5</mn></mrow></msubsup><mo>&gt;</mo><mn>0</mn><mo>.</mo><mn>20</mn></mrow></math></span> or for geometries with pronounced bends, an interface-capturing CFD model is essential to avoid overly conservative blockage estimates. These findings provide a clear basis for selecting and calibrating numerical tools in full-scale nuclear-plant safety assessments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114416"},"PeriodicalIF":2.1,"publicationDate":"2025-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144916552","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":"Safety margin evaluation of RSG-GAS during LOFA considering ageing-induced degradation of coast-down flow and recent decay heat measurement","authors":"Endiah Puji Hastuti ,&nbsp;Azizul Khakim ,&nbsp;Heri Suherkiman ,&nbsp;Topan Setiadipura ,&nbsp;Dwi Irwanto ,&nbsp;Abdul Waris","doi":"10.1016/j.nucengdes.2025.114413","DOIUrl":"10.1016/j.nucengdes.2025.114413","url":null,"abstract":"<div><div>This study evaluates the safety performance of the RSG-GAS research reactor during a Loss of Flow Accident (LOFA), with a particular focus on the effects of aging on primary pump coast-down behavior and post-shutdown decay heat. As the reactor has operated for over 38 years, mechanical degradation, especially in the pump flywheel, has led to a more rapid decline in coolant flow following a pump trip. To capture this effect, updated coast-down flow and decay heat profiles were experimentally measured and incorporated into LOFA simulations using the PARET/ANL code. The analysis compares reactor behavior under commissioning and current conditions at initial power levels of 15 MW and 30 MW. Results show that while the updated profiles slightly reduce safety margins, evident in lower burnout ratios and higher coolant temperatures, all key safety parameters remain within the reactor’s Operational Limits and Conditions (OLCs). The passive safety system, based on natural convection, effectively removes decay heat even with degraded pump inertia. These findings underscore the importance of incorporating real-time experimental data into safety evaluations and confirm that the RSG-GAS reactor continues to operate safely under ageing conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114413"},"PeriodicalIF":2.1,"publicationDate":"2025-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144916551","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":"Root cause analysis of a molten salt pump in FLUSTFA","authors":"Adam Burak ,&nbsp;Shuai Che ,&nbsp;Dan Barth ,&nbsp;Brandon Haugh ,&nbsp;David Holcomb ,&nbsp;Leonard Prokopchak ,&nbsp;Kevin Robb ,&nbsp;Vicente Rojas ,&nbsp;Sheng Zhang ,&nbsp;Xiaodong Sun","doi":"10.1016/j.nucengdes.2025.114365","DOIUrl":"10.1016/j.nucengdes.2025.114365","url":null,"abstract":"<div><div>The primary salt pump installed in the high-temperature FLUoride Salt Test Facility (FLUSTFA) was successfully operated for some time, but later ceased operation. To understand what occurred, a Root Cause Analysis (RCA) was performed. Steps taken to try to get the pump operational include adjusting the shaft position, increasing the heating power of the tape heaters on the pump volute, and manually rotating the pump shaft. While removing the insulation, corrosion was noted on the outside of the pump volute, and decolorization of the insulation and tape heaters was observed. Significant corrosion products were also observed in the pump itself and the piping connected to the pump. The nitrogen cover gas was maintained from before salt was introduced into the loop until the pump was dismounted and continues to be maintained even after the pump was removed. After considering probable scenarios, causes were assigned and corrective actions were developed to prevent those causes. Then, the RCA was presented to an advisory committee for review, the “Review Committee,” consisting of experts in large molten salt systems: Brandon Haugh, David Holcomb, Kevin Robb, and Vicente Rojas. The advisory committee provided comprehensive feedback, which have been incorporated into a revised RCA. Findings have then been summarized and reported in this publication.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114365"},"PeriodicalIF":2.1,"publicationDate":"2025-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144912625","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":"A PSO-based burst pressure equation of super duplex stainless steel pipelines at high temperatures","authors":"Shuqian Shen ,&nbsp;Yan Li ,&nbsp;Bao Zhang ,&nbsp;Yi Shuai ,&nbsp;Liyang Wu ,&nbsp;Xiaofu Chen ,&nbsp;Zhanfeng Chen","doi":"10.1016/j.nucengdes.2025.114427","DOIUrl":"10.1016/j.nucengdes.2025.114427","url":null,"abstract":"<div><div>Pipelines in nuclear, oil and gas, and chemical industries often operate under prolonged high-temperature and high-pressure conditions, which significantly degrade material strength and complicate burst pressure prediction. This study proposes a novel semi-empirical model to estimate the burst pressure of super duplex stainless steel (SDSS) pipelines under high temperature. Temperature-dependent mechanical properties were quantified through experiments and fitted into empirical equations. A new burst pressure equation was then developed, incorporating these thermal effects. A material-specific correction factor for SDSS was calibrated using Particle Swarm Optimization (PSO) based on finite element simulations. The proposed model was validated against both numerical and experimental results, showing a relative error below 4.2%. This method provides a reliable and practical tool for safety assessment of pipelines exposed to high-temperature service environments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114427"},"PeriodicalIF":2.1,"publicationDate":"2025-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144912627","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}