Nuclear Engineering and Design最新文献

{"title":"Inverse uncertainty Quantification in the Severe accident Domain: Application to Fission Product release","authors":"G. Tinfena ,&nbsp;M. Angelucci ,&nbsp;L. Sargentini ,&nbsp;S. Paci ,&nbsp;L.E. Herranz","doi":"10.1016/j.nucengdes.2025.113954","DOIUrl":"10.1016/j.nucengdes.2025.113954","url":null,"abstract":"<div><div>This paper presents a pioneering application of Inverse Uncertainty Quantification (IUQ) methodology, proposed by CSNI/WGAMA in the SAPIUM activity, within the Severe Accident (SA) domain. Framed under the SEAKNOT-EU project, such a systematic approach has been used for uncertainty quantification of the Fission Product Release (FPR), particularly the cesium fractional one, as estimated by the MELCOR code. More than 60 experiments have been analyzed and some of them selected based on their completeness and representativeness, as recommended by the adequacy analysis. By using the IUQ CIRCE method, the so called Revised CORSOR-Booth model uncertainty range has been estimated and proved consistent with the database. Finally, the uncertainty characterization has been validated against on-line experimental data recorded in PHEBUS-FPT1.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113954"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143580754","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Numerical modelling of thermal fatigue at pipe mixing points in nuclear power plants","authors":"Funke Dacosta-Salu ,&nbsp;Michael E Fitzpatrick ,&nbsp;Xiang Zhang ,&nbsp;Michael Roy ,&nbsp;James Jewkes","doi":"10.1016/j.nucengdes.2025.113978","DOIUrl":"10.1016/j.nucengdes.2025.113978","url":null,"abstract":"<div><div>In nuclear power plants, thermal fatigue can occur at pipe mixing points where hot and cold water combine, leading to failure at these critical locations. This study investigates the effect of temperature fluctuations on fatigue failure at critical locations and welded joints, which has received limited attention in previous research. A numerical approach was used, starting with highly-resolved unsteady conjugate heat transfer simulations to assess heat flux at the pipe wall. This was followed by structural analysis using the finite element method, and finally, a fatigue assessment to predict failure locations and estimate component lifespan. Temperature differences of 80 °C and 160 °C were investigated at the different weld locations. Results showed that at a temperature difference of 160 °C between the main pipe and the branch pipe, a full penetration butt weld would fail after 1462 to 19,119 h. A shorter failure time was observed at the stress concentration area upstream of the T-junction under the same conditions. These findings were applied to the well-documented 1998 failure at the Civaux1 plant in France, to help understand the potential causes of that failure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113978"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143580783","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Impact of community shielding on radiological risk following a hypothetical nuclear explosion","authors":"Osamong Gideon Akou ,&nbsp;Xuan Wang ,&nbsp;Shuhuan Liu ,&nbsp;Xinwei Liu ,&nbsp;Guanghui Su ,&nbsp;Ailing Zhang ,&nbsp;Junfang Zhang ,&nbsp;Minghua Lv ,&nbsp;Lei Huang ,&nbsp;Shanchao Yang","doi":"10.1016/j.nucengdes.2025.113986","DOIUrl":"10.1016/j.nucengdes.2025.113986","url":null,"abstract":"<div><div>The impact of a nuclear explosion is uncertain due to various factors such as the size of the explosion, the altitude at which it occurs, the weather conditions, and the surrounding terrain. Fission products released can have short-term or long-term effects, potentially leading to mortality. This study examined the effects of hypothetical nuclear detonations of 15 kT and 21 kT in a community using the HotSpot code. The main variables considered were shielding, wind speed, explosion power, time spent in the explosion zone, protection factors, and structure thickness. Existing structures significantly reduce external dose, prompt gamma, prompt neutron, and fallout. While a 21 kT explosion is more lethal than 15 kT, shielding is effective in attenuating radiation for both explosions. Proximity and duration of exposure to the explosion zone increase radiation effects. At higher velocities (4.5 m/s), the plume spreads more. High shielding effectiveness corresponds to low external dose levels. Concrete shielding is particularly more effective in attenuating and scattering radiation. Our study provides valuable guidance for decision-makers and the public on mitigation measures and sheltering prioritization in the event of a nuclear explosion.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113986"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143580784","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":"Identification and analysis of current signal characteristics of nuclear power pump units based on start-up process","authors":"Xiuli Wang ,&nbsp;Shenpeng Yang ,&nbsp;YiFan Zhi ,&nbsp;Qichao Xia ,&nbsp;Wei Xu ,&nbsp;Yuanyuan Zhao","doi":"10.1016/j.nucengdes.2025.113985","DOIUrl":"10.1016/j.nucengdes.2025.113985","url":null,"abstract":"<div><div>Monitoring and identification are the effective ways to ensure the safe and reliable operation of nuclear power pumps. In order to monitor and identify the operating status of nuclear power pumps with different impellers effectively, this paper uses six types of impellers for analysis, and thirteen current signals are collected for each impeller in the start-up process of the experiment. Signal preprocessing is performed by resampling, combined with Fast Fourier Transformation (FFT), Variable Mode Decomposition (VMD), Empirical Mode Decomposition (EMD), and VMD-EMD joint method for feature analysis, and it explores the relationship between characteristic signals and operating conditions, revealing the variation patterns of current signals in the start-up process of different impellers under different operating conditions. The results show that the amplitude of the current signal increases with the increasing of the flow rates and the impeller diameter in the start-up process. The EMD method has an error of less than 20% under low flow rates, but the range of the error is 36%-47% in the efficient working area, with a maximum of nearly 50%. The VMD method has an accuracy of about 80% in an efficient working area, but has a significant error at low flow rates, reaching up to 50%. The VMD-EMD method can solve the problem of low accuracy of results, and the accuracy is guaranteed to be above 80%. The accuracy is above 90% within the efficient range of 0.8Q₀-1.2Q₀. This method can effectively achieve fault diagnosis.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113985"},"PeriodicalIF":1.9,"publicationDate":"2025-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579667","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":"Operation characteristics analysis of supercritical CO2 reactor based on neutronics and thermal-hydraulics coupling","authors":"Yuchen Niu ,&nbsp;Dabin Sun ,&nbsp;Yuandong Zhang ,&nbsp;Lei Chen ,&nbsp;Minjun Peng ,&nbsp;Genglei Xia","doi":"10.1016/j.nucengdes.2025.113947","DOIUrl":"10.1016/j.nucengdes.2025.113947","url":null,"abstract":"<div><div>Small Modular Reactors (SMRs) have emerged as a focal point in future nuclear systems, owing to their high energy density, simple structure, flexible configuration, and extended endurance. The innovative nuclear system using supercritical carbon dioxide (SCO₂) Brayton cycle as the energy conversion system offers notable benefits such as high conversion efficiency, compact layout. Additionally, SCO₂ can achieve a greater density difference compared to water, potentially enabling an enhancement in inherent safety through natural circulation, which renders it stand out in SMRs. Nonetheless, the distinct thermo-physical properties of SCO₂ present challenges in understanding the multi-physics coupling characteristics of this system. Current operation characteristics analysis focus on individual physical fields, lacking the effective methods to calculate the inherent multi-physics coupling characteristics of SCO₂ reactors. This study introduced a neutronics and thermal-hydraulics coupling method for SCO₂ cooled reactors, based on the thermal-hydraulics analysis capabilities of RELAP5-SCO₂ and the three-dimensional neutron physics analysis capabilities of SIM-CORE. The coupling framework between the RELAP5-SCO₂ and the SIM-CORE was established. The time step control strategy, spatial grid mapping technique, and the method for transmitting coupling parameters between above two programs were described in detail. The proposed methods and developed models were validated by accurately representing the multi-physics coupling features of a 15MW_th SCO₂ direct cooled micro-reactor. Meanwhile, the multi-physics coupling characteristics under the typical multi-power operation conditions and representative accident conditions, including reactivity-initiated accident, loss of flow accident and loss of coolant accident, were analyzed. The three-dimensional core neutronics and thermal-hydraulics coupling program developed in this paper can enhance the understanding of the transient operating characteristics of SCO₂ gas-cooled reactors and the strong nuclear-thermal coupling effects in accident scenarios, providing theoretical support and technical assistance for the research and development and safety analysis of SCO₂ gas-cooled reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113947"},"PeriodicalIF":1.9,"publicationDate":"2025-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143580782","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":"Core loading optimization of Apsara-U research reactor using differential evolution algorithm","authors":"Y.S. Rana,&nbsp;Tej Singh","doi":"10.1016/j.nucengdes.2025.113980","DOIUrl":"10.1016/j.nucengdes.2025.113980","url":null,"abstract":"<div><div>A discrete differential evolution algorithm (DE) has been applied for optimizing core loading of Apsara-U research reactor. The objective is to maximize the core excess reactivity by restricting the maximum power of a fuel assembly to 173 kW. First, calculations were performed to verify the optimum values of mutation scale factor and crossover rate given in the literature. Subsequently, DE search was performed up to 500 generations with the population size of 10. It is found that the reactivity reaches its maximum value after 350 generations. The results show that, for a given core configuration, it is possible to obtain loading patterns which provide higher reactivity gain compared to the implemented pattern.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113980"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563196","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":"Dynamic dose equivalent rate estimation in dismantling: Physic-informed surrogate modeling","authors":"Pierre-Louis Antonsanti ,&nbsp;Geoffrey Daniel ,&nbsp;François Bachoc ,&nbsp;Cindy Le Loirec","doi":"10.1016/j.nucengdes.2025.113971","DOIUrl":"10.1016/j.nucengdes.2025.113971","url":null,"abstract":"<div><div>The estimation of dose equivalent rate plays a key role in the radiation protection strategy for decontamination and dismantling. In particular, real-time map estimation of dose equivalent rate provides a user interface for planning interventions on basic nuclear installations undergoing dismantling. Conventional approaches for this estimation rely either on Monte-Carlo simulation whose computational time is prohibitive for real-time applications or on deterministic approaches whose approximations deteriorate the precision of the estimation in complex configurations. This work focuses on the construction of surrogate models, designed to mitigate these limitations and to estimate in real-time the dose equivalent rate given a source position in a specific installation. These models are tuned using data from Monte Carlo simulations, and take advantage of additional information, called ”additional descriptors”. These descriptors embed the knowledge on the physical behavior of the particles in a specific simulated installation. Three surrogate models, the K nearest neighbors, XGBoost, and the Gaussian Process regression are compared, with and without the additional descriptors. They are evaluated on three configurations encountered in radiation protection. The results show that the physical information allows the surrogate models to adapt to new source positions in the geometry, and limits the size of the database needed to train the models.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113971"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563133","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":"Cross-Domain Few-Shot Anomaly Detection for equipment in nuclear power plants","authors":"Junjie He ,&nbsp;Sheng Zheng ,&nbsp;Shuang Yi ,&nbsp;Senquan Yang ,&nbsp;Zhihe Huan","doi":"10.1016/j.nucengdes.2025.113956","DOIUrl":"10.1016/j.nucengdes.2025.113956","url":null,"abstract":"<div><div>In Nuclear Power Plants (NPPs), operating data from equipment may shift due to changes in environmental conditions, device degradation, or component replacements. These shifts can impact the performance of data-driven monitoring models trained solely on source domain data, leading to increased false alarms and reducing both the effectiveness and reliability of the models. Furthermore, the amount of shifted data in real-time monitoring is limited and cannot meet the demands for deep learning model’s training process. To address the problems of Cross-Domain Few-Shot Anomaly Detection (CDFS-AD), we propose a Deep Temporal–Spatial Transfer Learning Network (DTSTLN). The proposed model leverages an improved transformer model to achieve temporal–spatial feature extraction and reconstruction of input operating data. And Maximum Mean Discrepancy (MMD) based loss function is utilized to achieve domain adaptation, enabling knowledge transfer and effective training with limited data. Comparative experiments on real operating data from the reactor coolant pump in NPPs demonstrate the effectiveness of DTSTLN in monitoring shifted data, as evidenced by higher F1-scores and lower False Alarm Rates (FARs) compared to other baseline methods, highlighting its potential for anomaly detection of NPP equipment in real scenarios.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113956"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563195","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 new cross-scale model for leakage-rate prediction of metal-to-metal seals under high-pressure conditions","authors":"Weixi Zhang ,&nbsp;Jinhui Wang ,&nbsp;Xiaoming Huang ,&nbsp;Guoliang Xu ,&nbsp;Du Zhou","doi":"10.1016/j.nucengdes.2025.113981","DOIUrl":"10.1016/j.nucengdes.2025.113981","url":null,"abstract":"<div><div>In order to better reveal the leakage mechanism of metal-to-metal valves under high-pressure condition in nuclear power plants, a cross-scale prediction method is proposed, which integrates various simulation techniques such as mesoscopic flow, mesoscopic contact, and macroscopic contact. On this basis, the flow resistance network analysis method is introduced to better evaluate the influence of high-pressure liquid penetration on the leakage characteristics of metal-to-metal valves. In this manner, the model can quantitatively evaluate the impact of various parameters on the leakage rate, such as surface morphology, geometric structure, sealing load, and fluid pressure. A leakage-rate test device of multi parameter metal-to-metal seals is designed and constructed. The effectiveness of the theoretical prediction method is verified based on the experimental results. Subsequent simulation studies have revealed that as the medium pressure increases or the sealing force decreases, the distribution of contact stress on the sealing surface exhibits three distinct patterns: unpenetrated, near-penetrated, and penetrated. The leakage rate is highly sensitive to changes of sealing force in the penetrated pattern, while the leakage rate is highly affected by surface morphology in the unpenetrated pattern. Curves of required contact stresses and fluid pressures were calculated and compared for different contact widths, using the maximum allowable leakage rate (<em>Q_lim</em> = 1.2 g/min) as an indicator. An abnormal increase in the slope of the curves (from 0.27 to 4.95) was found for high pressures (<em>P_fluid</em> &gt; 5 MPa), high roughness (0.3 μm) and small contact widths (4 mm), suggesting that the sealing performance is dominated by a combination of micro-flow and macro-contact. Reverse penetration can greatly reduce the sensitivity of leakage rate to sealing force. The research in the article will permit improved design techniques that significantly improve the sealing performance of metal-to-metal seals.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113981"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143551481","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":"Startup characteristics of space nuclear power system with multiple Brayton loops from an electric perspective","authors":"Chuping Yang,&nbsp;Wenkui Ma,&nbsp;Yinan Geng,&nbsp;Xiaoyong Yang,&nbsp;Jie Wang","doi":"10.1016/j.nucengdes.2025.113959","DOIUrl":"10.1016/j.nucengdes.2025.113959","url":null,"abstract":"<div><div>The closed Brayton cycle (CBC) is a promising power conversion unit (PCU) to convert heat from the reactor to electric power for high-power space nuclear power system (SNPS) because of its high efficiency, small volume, light weight and stable operation. To ensure system redundancy and eliminate the gyrostatic effect, the SNPS usually uses multiple Brayton loops (SNPS-MBL), with each Brayton loop sharing one reactor. For the SNPS-MBL, the startup characteristic is an essential part. In this study, the simulation model of the startup was established. The startup processes of simultaneous startup and sequential startup for SNPS with dual Brayton loops (SNPS-DBL) were compared and analyzed from the electric perspective, and the effect of mechanical load torque on the startup process was further analyzed. This study provides a reference for the starting process of the SNPS from an electric perspective.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113959"},"PeriodicalIF":1.9,"publicationDate":"2025-03-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143551480","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}