Nuclear Engineering and Design最新文献

{"title":"Toward development of a low-temperature failure envelope of cases for high-burnup RIAs under PWR operational conditions","authors":"L. Aldeia Machado ,&nbsp;K. Nantes ,&nbsp;E. Merzari ,&nbsp;L. Charlot ,&nbsp;A. Motta ,&nbsp;W. Walters","doi":"10.1016/j.nucengdes.2024.113642","DOIUrl":"10.1016/j.nucengdes.2024.113642","url":null,"abstract":"<div><div>A Reactivity-Initiated Accident (RIA) is a design-basis accident that occurs when the reactor loses one of its control rods. A reactivity insertion will follow such events, drastically increasing the fuel pellet’s temperature and volume due to thermal expansion. The fuel pellet and the cladding will interact mechanically, which could lead to cladding failure. This work presents the development of cases where low-temperature failures are more likely to happen for high-burnup fuels under PWR operational conditions. A coupled computational model between the nuclear fuel performance code BISON, MOOSE’s Thermal-Hydraulic Module (MOOSE-THM), and MOOSE’s Stochastic Tools Modules (MOOSE-STM) was created to study the thermal-hydraulic behavior of a high-burnup fuel rodlet during the first stages of an RIA transient, allowing us to identify three scenarios: the system reached CHF, leading to high-temperature failure, the system failed due to PCMI, or the system survived the whole transient without failing. To address these scenarios, we performed a sensitivity analysis with more than 140,000 model replicates through MOOSE-STM varying parameters such as the power pulse width, power pulse total energy deposition, hydride rim thickness, coolant inlet temperature, and coolant inlet mass flux. We also compared two PCMI failure criteria in our analysis. Our results suggest that the hydride rim thickness and the power pulse width will be the key parameters impacting the failure type our system would undergo during the power transient. Using the data from our simulations, we constructed two failure maps, one for each PCMI failure criterion, showing how the failure type is affected by each parameter considered in the sensitivity analysis. We also provided a closed-form expression for the boundary between the PCMI and CHF failure types as a function of the hydride rim thickness and power pulse width.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113642"},"PeriodicalIF":1.9,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552531","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":"Sub-channel analysis of the influence of the ATF cladding corrosion on thermal hydraulic behaviors","authors":"Mingdong Kai ,&nbsp;Jiejin Cai","doi":"10.1016/j.nucengdes.2024.113668","DOIUrl":"10.1016/j.nucengdes.2024.113668","url":null,"abstract":"<div><div>Accident-tolerant fuel (ATF) enhances the accident tolerance of the fuels by improving its thermal properties and antioxidant radiation performance, thereby enabling the reactor to withstand severe accidents for a long time. This article applies and improves heat transfer and CHF models in the COBRA-EN by considering the impact of cladding surface corrosion on critical heat transfer between coolant and fuel rods. We conduct detailed and critical validation of the model constructed in this paper based on two benchmark experiments. We apply this model to study the thermal–hydraulic behaviors of ATFs under accident conditions. We obtain parameters such as the maximum fuel centerline temperature (MFCT), the maximum cladding surface temperature (MCT), the minimal departure from nucleate boiling ratio (MDNBR), the critical heat flux (CHF), and the average void fraction (AVF) for different ATFs. The results indicate that under most transient operating conditions, cladding corrosion delays the soaring time of the MFCT and MCT, and generally enhances CHF, with an average enhancement amplitude of over 0.15 MW/m². At the same time, due to the effect of cladding corrosion, the MDNBR of the reactor has also been improved, which mitigates the impact of the accident to some extent.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113668"},"PeriodicalIF":1.9,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552533","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":"New insight into fatigue life of modified 9Cr-1Mo steel in liquid lead–bismuth environment and life prediction considering environmental factors","authors":"Shouwen Shi ,&nbsp;Wei Huang ,&nbsp;Gaoyuan Xie ,&nbsp;Weibin Li ,&nbsp;Longyi Yang ,&nbsp;Qiang Lin ,&nbsp;Gang Chen ,&nbsp;Xu Chen","doi":"10.1016/j.nucengdes.2024.113648","DOIUrl":"10.1016/j.nucengdes.2024.113648","url":null,"abstract":"<div><div>The fatigue life of modified 9Cr-1Mo steel in liquid lead bismuth eutectic (LBE) at different strain amplitudes, temperatures and oxygen concentrations are analyzed. A liquid metal embrittlement (LME) factor of plastic strain is proposed to account for the reduced fatigue life induced by LME effect, which is also found to correlate well with tensile elongation in LBE. In low oxygen content LBE, the LME effect is influenced by temperature instead of plastic strain amplitude. While in high oxygen content LBE, the plastic LME factor is found to decrease exponentially with increasing plastic strain amplitude. Based on these findings, a fatigue life prediction model is proposed taking into account of different environmental influencing factors. In total, 86 data points are used with 70 % data points for independent validation only. Regardless of the discrepancy in fatigue life from different sources, good prediction results are still achieved with 98 % data points fall within ± 3 error band and 75 % data points fall within ± 2 error band.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113648"},"PeriodicalIF":1.9,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552534","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":"Analysis of the experimental tests performed at NACIE-UP facility through a novel CFX-RELAP5 codes coupling","authors":"T. Del Moro ,&nbsp;P. Cioli Puviani ,&nbsp;B. Gonfiotti ,&nbsp;I. Di Piazza ,&nbsp;D. Martelli ,&nbsp;C. Ciurluini ,&nbsp;F. Giannetti ,&nbsp;R. Zanino ,&nbsp;M. Tarantino","doi":"10.1016/j.nucengdes.2024.113676","DOIUrl":"10.1016/j.nucengdes.2024.113676","url":null,"abstract":"<div><div>The design and safety assessment of Lead-cooled Fast Reactors (LFRs), being one of the Generation IV technologies, must be supported by extensive experimental campaigns. Such activities are necessary to completely understand the physical phenomena involved in such reactors, as well as to properly develop new numerical tools or validate the pre-existent ones. From the experimental point of view, ENEA Research Center of Brasimone is one of the most active institutions, thanks to its experimental platforms and know-how maturated since the early 2000s. From the numerical point of view, Computational Fluid Dynamics (CFD) codes are the most suitable ones to analyze some phenomena expected in a Heavy Liquid Metal (HLM)-cooled reactor, such as the complex 3D phenomena occurring within the pools or the core fuel assemblies. In addition, the fluid thermal conduction, usually neglected in a System Thermal-Hydraulic (STH) code, can assume a significant importance in some transient scenarios, e.g., loss of flow accidents with transition from forced to natural circulation. However, the safety analysis of the LFRs should still rely on the use of STH codes because of their lower computational cost compared to the CFD codes, also considering the high number of transient evolutions to be analyzed for the purpose of the reactor licensing. At ENEA Brasimone, a novel coupling approach has been developed to couple the CFD code Ansys CFX with the STH code RELAP5/Mod3.3. The coupled tool aims at exploiting the advantages of the two families of codes. It adopts a multi-scale approach to simulate in detail some circuit components while performing system-level analysis, so as to keep an acceptable computational time. The coupling technique is based on ad-hoc user routines written in FORTRAN and implemented in Ansys CFX, which acts as the master code. The user routines take care of time step management, data exchange, RELAP5 execution, and error checking. The goal of this paper is to assess the simulation capabilities of the coupled tool by reproducing a forced-to-natural-circulation transition test, carried out at the NACIE-UP facility, with LBE as working fluid. The work has been realized in the framework of the IAEA Coordinate Research Project-I31038, named “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop”.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113676"},"PeriodicalIF":1.9,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552532","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":"Crack path analysis of spent nuclear fuel cladding using the strain energy-based Dijkstra algorithm","authors":"Jee A Baik,&nbsp;Jung Jin Kim","doi":"10.1016/j.nucengdes.2024.113661","DOIUrl":"10.1016/j.nucengdes.2024.113661","url":null,"abstract":"<div><div>The integrity of spent fuel cladding is crucial for preventing the release of radioactive materials, which pose significant risks to public safety and the environment. However, accurately predicting cracks in cladding tubes remains a challenge. This study proposes a novel method for predicting crack paths in spent nuclear fuel cladding tubes using the Dijkstra algorithm, based on strain energy. In this method, cladding images are segmented into cladding and hydride pixels, followed by a finite element analysis to calculate the strain energy. The Dijkstra algorithm utilizes this strain energy data from hydrides to predict crack paths in areas with low resistance to loading. The predicted path exhibited an accuracy of 92.78 % with respect to the initiation point of the actual crack path and was located within 200 μm of the actual crack path. The proposed method demonstrates a higher similarity to the actual crack path than conventional image-based methods. These results suggest that the safety assessment of spent nuclear fuel can be enhanced, enabling the development of effective management strategies for spent nuclear fuel.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113661"},"PeriodicalIF":1.9,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552681","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":"Simulating the time-dependent evolution of Alkali-Silica Reaction (ASR) strains in concrete","authors":"Gunay Gina Aliyeva ,&nbsp;Yann Le Pape ,&nbsp;Abhinav Gupta","doi":"10.1016/j.nucengdes.2024.113658","DOIUrl":"10.1016/j.nucengdes.2024.113658","url":null,"abstract":"<div><div>Alkali-Silica Reaction (ASR) affects the resiliency of concrete structures by initiation of cracking in concrete which in turn leads to deterioration. There has been an increasing demand to understand the ASR-induced expansion and degradation in concrete. Continued safe operation of concrete structures requires an assessment of ASR-induced expansion and degradation. This paper attempts to understand the time-dependent evolution of ASR-induced expansion and degradation in concrete structures. A novel approach is proposed to simulate the ASR-induced expansion and degradation in concrete that is based on coupling the ASR-induced strains with the mechanical strains using a time-dependent piecewise evolution process at each instance of time. Data from an experimental study is used to develop the proposed approach. It is shown that the proposed approach is able to simulate the ASR-induced expansion and degradation in concrete reasonably well.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113658"},"PeriodicalIF":1.9,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552682","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":"Open-set recognition based on the combination of deep learning and hypothesis testing for detecting unknown nuclear faults","authors":"Wei Pan ,&nbsp;Jihong Shen ,&nbsp;Bo Wang ,&nbsp;Shujuan Wang ,&nbsp;Zhanhao Sun","doi":"10.1016/j.nucengdes.2024.113654","DOIUrl":"10.1016/j.nucengdes.2024.113654","url":null,"abstract":"<div><div>Most current fault diagnosis techniques for nuclear systems mainly rely on the closed-set assumption, which restricts the diagnosis model to select from a set of pre-established known fault classes. However, the nuclear system is a dynamic open system, and unknown faults that have never been seen can occur at any time. Therefore, it is very meaningful to design a diagnosis model that can recognize both known and unknown faults. This paper proposes a fault diagnosis method for open-set scenarios. Specifically, a modified loss function is used to train a convolutional neural network (CNN) to learn more compact feature representations of known classes. The features output by the last fully connected layer of the CNN are taken as the scores belonging to each known class, and a calibration model based on extreme value theory (EVT) is introduced to calibrate the scores. In addition, hypothesis testing is introduced for statistical inference. The threshold is determined according to the confidence level to distinguish the known faults from the unknown faults. Experiments conducted on two sets of nuclear system faults simulation data demonstrate that the proposed model not only identifies more unknown faults without compromising the accuracy of known fault classification but also selects more appropriate thresholds for different datasets, thereby enhancing the model’s generalization capability. Furthermore, experiments under varying degrees of openness also prove that our model exhibits higher robustness across different levels of openness.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113654"},"PeriodicalIF":1.9,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552683","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 against CABRI-E7 experiment by coupling fuel-pin failure module with FRTAC","authors":"Yi Lei ,&nbsp;Bin Zhang ,&nbsp;Siqi Feng ,&nbsp;Hao Yang ,&nbsp;Shaowei Tang ,&nbsp;Lin Sun","doi":"10.1016/j.nucengdes.2024.113651","DOIUrl":"10.1016/j.nucengdes.2024.113651","url":null,"abstract":"<div><div>Fuel-pin failure hold considerable importance in safety evaluations of sodium-fast reactors (SFRs) as fuel swelling and cladding rupture are key phenomena in the early stages of core disruptive accidents (CDAs). For transients leading to pin failure, the failure modes and initial fuel disruption depend partly on pre-transient irradiation effects, such as fission-gas retention and release, fuel swelling, cladding deformation, and central void formation. With the increasingly stringent requirements on safety analysis, it is necessary to accurately evaluate the thermo-mechanical degradation of the fuel-pin resulting from pre-transient irradiation. Therefore, this study developed a fuel-pin failure module based on mechanistic models of pre-transient fuel-pin characterization and proposed an innovative approach by coupling this module with the self-developed Fast Reactor Transient Analysis Code (FRTAC). The results of the numerical simulation against the CABRI-E7 test are presented and discussed in this paper. The expected heat transfer mechanism between fuel and cladding was reproduced by the simulation, and the temperature distribution of the fuel pin agreed well with other reference analysis codes. Additionally, analyses based on elastoplastic mechanics theory and biaxial stress rupture criteria were conducted, with a specific focus on the thermal and mechanical failure of the fuel-pin. The overall code assessment indicated that the prediction error was within an acceptable range, demonstrating that the module’s reliability and its applicability to safety analyses of oxide fuel in CDAs of SFRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113651"},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535541","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":"Round-robin analysis of highly depleted lithium for Generation IV nuclear reactor applications","authors":"Sean R. Scott ,&nbsp;Johnny Williams ,&nbsp;Sara Mastromarino ,&nbsp;Norbert Gajos ,&nbsp;Christian Berry ,&nbsp;Ian Anderson ,&nbsp;Steven Shen ,&nbsp;Trent R. Graham ,&nbsp;Cole Hexel ,&nbsp;Josh Wimpenny ,&nbsp;Jacob Brookhart ,&nbsp;Alan Kruizenga","doi":"10.1016/j.nucengdes.2024.113664","DOIUrl":"10.1016/j.nucengdes.2024.113664","url":null,"abstract":"<div><div>Lithium reference materials containing unnaturally high abundances of ⁷Li are not currently available, which poses quality control problems for highly depleted lithium materials (i.e., depleted in ⁶Li) required for Generation IV nuclear reactors. This study presents an interlaboratory comparison of a lithium carbonate (NIST SRM924a) containing nominally natural isotopic abundances (∼92.4 % Li-7) and a highly depleted lithium hydroxide material (∼99.95 % Li-7). The natural lithium isotope abundances of NIST SRM924a are confirmed, and the ⁶Li/⁷Li ratio of the lithium hydroxide ranged from 0.000399 to 0.000436 with an average of 0.000428 ± 0.000023 (2SD, n = 9). Going forward this material can be used as quality control for analytical work involving highly depleted lithium.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113664"},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535542","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":"CFD modeling methods applied to a lead-cooled fast reactor: A parametric study on conjugate heat transfer and thermal boundary conditions","authors":"Ivan K. Umezu ,&nbsp;Dario M. Godino ,&nbsp;Damián E. Ramajo ,&nbsp;Claubia Pereira ,&nbsp;Antonella L. Costa","doi":"10.1016/j.nucengdes.2024.113649","DOIUrl":"10.1016/j.nucengdes.2024.113649","url":null,"abstract":"<div><div>Given the ever-increasing global demand for energy and the need to reduce greenhouse gas emissions, small modular reactors (SMRs), have emerged as potential options for increasing the contribution of nuclear energy, offering lower costs and faster deployment compared to traditional nuclear projects. In the context of this technological development, safety studies have become a priority, particularly for licensing new-generation systems such as metal-cooled fast reactors. This work models the steady-state operation of the lead-cooled SMR SEALER Arctic using Computational Fluid Dynamics. The entire primary circuit of the SEALER is modeled; the core is represented as a combination of porous media and heat sources, the pumps are represented as recirculating boundary conditions to account for momentum sources, and the steam generators are represented as porous media coupled with a temperature-dependent heat sink function. The main objective of this study is to simulate the SEALER under steady-state condition, while also accounting for the effects of heat conduction through its solid regions, and heat losses on the reactor vessel wall to the environment. For the former, the reactor is modeled with and without conductive solids and surfaces, using a conjugate heat transfer model. For the latter, natural convection and radiation heat transfer considerations are included as boundary conditions, and a parametric study is carried out with a range of external temperatures, and their effects on fuel and coolant temperatures are also discussed. Despite significant differences in local temperatures near the vessel walls, the impact on the peak fuel temperature and the average coolant temperature was less noticeable. Ultimately, the general operating parameters of the steady-state reactor design were verified, which is the first step before using the current model to evaluate fast transients and postulated events, where the thermal inertia of the solids and additional heat losses could play a crucial role on determining the system’s response to rapid temperature changes.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113649"},"PeriodicalIF":1.9,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535538","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}