Nuclear Engineering and Design最新文献

{"title":"Numerical simulation of seismic response of spent fuel pool considering structure-fluid-structure interaction","authors":"Tsvetan Genov,&nbsp;Joao Travanca,&nbsp;Stoyan Andreev,&nbsp;Anton Andonov","doi":"10.1016/j.nucengdes.2025.113892","DOIUrl":"10.1016/j.nucengdes.2025.113892","url":null,"abstract":"<div><div>The objective of the paper is to assess approaches to include the effect of fluid–structure along with equipment-fluid–structure interaction by evaluating the response of a typical spent fuel pool found in Westinghouse type Pressurized water reactor auxiliary building to earthquake loading, as well as to derive the seismic demand in critical structural members for the needs of a seismic fragility analysis. Different approaches are presented to account for fluid–structure interaction during the dynamic nature of seismic loading. All structural key aspects considered relevant are included in the analyses: the concrete box structure, the outlining stainless-steel liner, its connections, the submerged rack’s structure, and the containing fluid.</div><div>Three numerical models were developed and tested with the finite element code LS-Dyna with different implementations considering the fluid–structure interaction to demonstrate the robustness of the structure against earthquakes. Mechanical analogue adopted in codes with spring-mass system representation of the fluid attached to the structure and explicit modelling of the domain representing the water considering the two-way interaction between structure and fluid.</div><div>It is of great importance to include all the effects of fluid–structure interaction, as they increase the response and load of the structure. Both approaches considered for fluid representation were found to capture accurately the global response of the structure. The lattice structure of the racks exerts extra damping in the fluid motion and the pressure loading on the tank walls in near vicinity during earthquake excitation. The explicit modelling approach for the fluid elements could possibly result in lowering the racks critical displacements.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113892"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169293","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Investigation on the effect of operational parameters in a microreactor system on the morphology and size distribution of thorium oxalate","authors":"P. Zaheri,&nbsp;S. Ammari Allahyari,&nbsp;A. charkhi","doi":"10.1016/j.nucengdes.2024.113724","DOIUrl":"10.1016/j.nucengdes.2024.113724","url":null,"abstract":"<div><div>Thorium oxide has recently garnered attention as a nuclear fuel due to the scarcity of uranium resources and the abundance of thorium resources, as well as its favorable thermal and neutronic properties. One of the most desirable characteristics of thorium nuclear fuels is their thermal properties, which are affected by the size distribution and morphology of thorium oxide particles. Thorium nitrate is the most common method for producing thorium oxide. Since no purification processes occur in the production of thorium oxide from oxalates, particle size and shape control are crucial. In this paper, the synthesis of thorium oxalate particles is carried out in a microreactor system to control the parameters of the precipitates. The main parameters that influence the efficiency, size distribution, and morphology of thorium oxalate are the concentration ratio of oxalic acid to thorium nitrate solution, as well as the flow rate ratio of these feed materials. Results show that at lower flow rate ratios of thorium nitrate to oxalic acid solution and higher concentration ratios of acid to thorium nitrate solution, a uniform particle size distribution and smaller particles are obtained, which are suitable for further calcination to prepare high-density and small grain size pellets.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113724"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167567","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":"Finite element formulations of the semi-analytical time-domain model for flow-induced vibration of tube bundles in steam generators","authors":"Pan Sun ,&nbsp;Junzhe Shen ,&nbsp;Xielin Zhao ,&nbsp;Jinxiong Zhou","doi":"10.1016/j.nucengdes.2024.113769","DOIUrl":"10.1016/j.nucengdes.2024.113769","url":null,"abstract":"<div><div>This paper describes the finite element method (FEM) formulations for flow-induced vibration of steam generator tube bundles, based on a semi-analytical time domain (SATD) model proposed by us recently (Sun et al., Appl. Math. Modell. 132, 2024, 252-273). The SATD model is featured by explicit fluid force expressions in terms of multiple integrals with time delays. The new SATD model allows for a unified treatment of both time domain and frequency domain fluidelastic instability (FEI) analysis of a tube due subjected to crossflow. Specifically, this study adds to the body of knowledge through two contributions: (1) This paper presents the details of FEM formulations on this newly formulated SATD theory, in particular the discretization of fluid forces in both lift and drag directions, which are crucial for FEM implementation but missing in the literature; (2) This study presents the formulations based on FEM discretization for frequency domain FEI analysis of tubes, and provides very the first reference on FEI stability analysis of U-tubes. Our FEM results agree well with reported experimental data and numerical results. The numerical examples presented here, including single-span straight tubes subjected to uniform flow and multi-span U-tubes subjected to nonuniform flow, not only demonstrate the merit of the theory for academic research, but also exhibit the potential of the FEM code for realistic engineering applications. Our efforts provide useful and powerful numerical tools for flow-induced vibration analysis of tube bundles in steam generators and other heat exchangers.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113769"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167755","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":"Impact assessment of the integration of a generic PEM electrolyser facility into Rivne nuclear station","authors":"Eduard Diaz-Pescador,&nbsp;Marco Viebach,&nbsp;Florian Gamaleja,&nbsp;Wolfgang Lippmann,&nbsp;Antonio Hurtado","doi":"10.1016/j.nucengdes.2024.113767","DOIUrl":"10.1016/j.nucengdes.2024.113767","url":null,"abstract":"<div><div>This manuscript presents the outcomes from the impact assessment applied to the onsite integration of a 40 MW hydrogen production plant (HPP) into the protected area of Rivne nuclear power plant (NPP) in Ukraine. The scope of this work is limited to frequency estimation and blast loading characterization applied to three cases of study, representative of the worst-case scenario at different HPP locations. In the framework of the Euratom NPHyCo project, the NPP operator has proposed integration locations and provided corresponding distances to nearby structures with corresponding fragility criteria. The presented study assumes HPP operation at full capacity with allocated NPP electricity as a feedstock.</div><div>Hazardous points within the HPP are identified through HAZID (HAZard IDentification) methodology, which is assisted with system reliability analysis based on component failure rates. The identification of accident sequences in the HPP and frequency estimation is accomplished by means of event tree analysis (ETA). The study identifies as the worst-case scenario within the HPP a hydrogen explosion due to the destructive potential over far-range distances via shock wave propagation. In the electrolyser system, an unintended release of hydrogen may trigger a vapour cloud explosion (VCE), whereas in the adjacent hydrogen buffer tank a physical explosion may trigger a shock wave with subsequent projectile generation. The results show that the highest frequency of a hydrogen VCE is in the order of <span><math><mrow><msup><mrow><mn>10</mn></mrow><mrow><mo>-</mo><mn>7</mn></mrow></msup></mrow></math></span> per year following a 1% leak in the separator vessel upstream flange within the gas processing area.</div><div>The blast load characterization shows that a hydrogen VCE within the electrolyser facility would not exceed the 10 kPa fragility criterion of nearby safety-related structures. Contrarily, a physical explosion in the 30 kg buffer tank may lead to projectile generation with the potential to reach the standby diesel generators, cooling towers, and turbine building at a distance up to 500 m. The vessel fragments are deemed as soft missiles with high deformability upon impact. Nevertheless, means of protection are proposed to reduce the risk of the coupled facility. The outcomes of the study are meant as recommendations for a subsequent comprehensive safety assessment by the NPP operator.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113767"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168325","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":"A comprehensive analysis of neutronic properties of annular dispersed particle fuel","authors":"Song Li ,&nbsp;Lei liu ,&nbsp;Yongfa Zhang ,&nbsp;Jianli Hao ,&nbsp;Jiannan Li ,&nbsp;Qi Cai ,&nbsp;Qian Zhang","doi":"10.1016/j.nucengdes.2024.113786","DOIUrl":"10.1016/j.nucengdes.2024.113786","url":null,"abstract":"<div><div>This work made a comprehensive analysis of the neutronic performance of annular dispersed particle fuel (ADF), and the results were compared with annular ceramic fuel (ACF), cylindrical dispersed particle fuel (CDF), and cylindrical ceramic fuel (CCF). The four types of cells are simulated by in-house code, and then the effective multiplication factor, neutron flux distribution, depletion characteristics, nuclides composition variation, and temperature coefficient of each fuel rod are compared and analyzed. Moreover, the characteristics of fuel assembly consisting of the four types of fuels are analyzed, including the effective multiplication factor, normalized pin power distribution, etc. The calculation results show that ADF has obvious advantages in all the indicators mentioned above. The calculation and analysis conducted in this paper could compare the advantages of the new annular dispersed particle fuel rod over the traditional fuel rod, which provides a certain reference for the application of the new fuel rod in engineering.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113786"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168371","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":"Investigations on measurement and methodology for dielectric constant of solvents used for actinide separations in nuclear fuel cycle","authors":"Ajay Kumar Keshari ,&nbsp;J. Prabhakar Rao ,&nbsp;C.V.S. Brahmananda Rao","doi":"10.1016/j.nucengdes.2024.113734","DOIUrl":"10.1016/j.nucengdes.2024.113734","url":null,"abstract":"<div><div>PUREX (Plutonium uranium reduction extraction) is a versatile reprocessing method being employed in nuclear fuel cycle to separate U/Pu from the spent fuel present in 4 M nitric acid medium. The process is carried out by using 1.1 M tributyl phosphate (TBP)/n-DD (dodecane) as the extractant. High aqueous solubility and radiological and chemical degradation followed by retention of actinides are major disadvantages. However, to overcome certain demerits encountered by TBP and other alkyl phosphate-based extractants, organic phosphonate-based extractants have been developed and used to extract actinide.</div><div>Various organophosphorous based solvents are synthesized to develop better extractants for separation and recovery of actinides by solvent extraction in nuclear reactor/allied facilities. The polarity of the organic phase in the solvent extraction is a key parameter, which determines the amount of metal that can be loaded in the organic phase. The dielectric constant is an admissible ability to dissolve common ionic compounds. Typically, the dielectric constant of a solvent is evaluated for polarity. Higher dielectric constant is a measure of higher polarity, and greater is the capability of the solvent to hold the steady charges. The instrumentation setup is devised and developed for the measurement of the dielectric constant of solvents in a given medium The challenge in the setup is developing a dielectric sensor and instrumentation, that is compatible with sophisticated solvents possessing wide dynamic response with high accuracy, stability and precision. The investigations of the sensor, measurement methodology and estimation of the dielectric constant with various known and unknown solvents for the separation and recovery of actinides in the nuclear fuel cycle and nuclear waste treatment of the fast breeder test reactor were carried out and studied in detail. The dielectric constant of various standard solvents lying in the range 1–40 was measured and validated from literature values with an accuracy of ±1 %. Subsequently, the measurement of dielectric constant was carried out over different solvents such as Di-<em>iso</em>-amyl-<em>iso</em>-amyl phosphonate (DiAiAP), Di-<em>iso</em>-amyl-butyl phosphonate (DiABP), Di-<em>sec</em>-amyl-hydrogen phosphonate (DsAHP) and Di-<em>sec</em>-butyl-hydrogen phosphonate (DsBHP) synthesized in our laboratory.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113734"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168735","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 machine learning and CFD based approach for fouling rapid prediction in shell-and-tube heat exchanger","authors":"Shaopeng He,&nbsp;Yibo Ye,&nbsp;Mingjun Wang,&nbsp;Jing Zhang,&nbsp;Wenxi Tian,&nbsp;Suizheng Qiu,&nbsp;G.H. Su","doi":"10.1016/j.nucengdes.2024.113759","DOIUrl":"10.1016/j.nucengdes.2024.113759","url":null,"abstract":"<div><div>Rapid prediction of fouling and heat transfer performance over the full life cycle of an industrial heat exchanger is critical. Establishing a rapid-prediction model of the heat exchanger to realize full life cycle monitoring is an effective approach to assistant its operation, decontamination, and maintenance. With the advancement of machine learning, rapid prediction of heat exchanger fouling level used deep learning methods has become a research hotspot. Based on machine learning and CFD methods, a rapid fouling prediction surrogate model for the industrial heat exchanger was proposed on the example of steam generators in nuclear power plants. The mathematical model of fouling layer thermal resistance was developed. The 3D numerical simulation under different fouling levels was carried out. The high fidelity CFD simulation database was built, and four deep learning models (BPNN, PSO-BPNN, CNN, RBFNN) were adopted. The fouling thermal resistance of SG could be predicted rapidly according to operating parameters. The root-mean-square error of the four neural networks is less than 10⁻⁷ K/W. BPNN with PSO algorithm achieves the best balance between calculation time and prediction accuracy. The anti-noise performance of the prediction surrogate model was evaluated at different noise level of actual operating parameters. When the noise level is 5 %, predicted <em>R²</em> remains at 0.7615 and the mean relative error is still less than 15 %. The low-cost and fast prediction surrogate model developed in this paper can provide an effective reference for the maintenance and decontamination of industrial heat exchanger.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113759"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167513","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 gas entrainment evaluation model based on distribution of pressure along vortex center line – Application to a gas entrainment experiment with traveling vortices in an open water channel flow –","authors":"Kentaro Matsushita ,&nbsp;Toshiki Ezure ,&nbsp;Masaaki Tanaka ,&nbsp;Yasutomo Imai ,&nbsp;Tatsuya Fujisaki ,&nbsp;Takaaki Sakai","doi":"10.1016/j.nucengdes.2024.113785","DOIUrl":"10.1016/j.nucengdes.2024.113785","url":null,"abstract":"<div><div>Establishing an evaluation method for the gas entrainment (GE) of argon cover gas due to surface vortices is required from the perspective of safety design of sodium-cooled fast reactors (SFRs). An in-house evaluation tool for GE evaluation named StreamViewer has been developed. In previous studies, an evaluation model (original model) assuming the Burgers stretch vortex was implemented in the StreamViewer to evaluate the vortex dimple depth (gas core length). It was based on the calculation results of the pressure decrease at the vortex center point at the free surface. Since the conservativeness of the StreamViewer evaluation result with the original model has been proposed in a specific condition for a pool-type SFR, a modified evaluation model on the pressure distribution along the vortex center line (PVL model) was proposed to identify the vortex center lines as evaluation targets by connecting continuous vortex center points from the suction port to the surface using all vortex center points in an evaluation area. In the PVL model, each gas core length was evaluated based on the balance between the hydrostatic pressure and the pressure decrease distribution along the vortex center line using the three-dimensional results of computational fluid dynamics analysis. The applicability of the PVL model was confirmed by performing three-dimensional numerical analyses for the experiments where a rectangular thin plate induced unsteady traveling vortices in the open channel flow. Consequently, the GE evaluation using StreamViewer with the PVL model for the numerical analysis results could reproduce the relation between the inlet flow velocity and the gas core length, in other words, the elongation behavior of the gas core length with increased inlet velocity, in the unsteady vortex flow experiments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113785"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167560","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":"Model-based systems engineering adoption in the U.S. Nuclear industry","authors":"Jonathan K. Corrado","doi":"10.1016/j.nucengdes.2024.113752","DOIUrl":"10.1016/j.nucengdes.2024.113752","url":null,"abstract":"<div><div>This paper examines research covering applications of Model-Based Systems Engineering (MBSE) and the field of nuclear energy in the U.S. to gain a better understanding of how to better posture the industry for expansion while maintaining a high margin of safety. It provides an analysis of current research to identify the benefits and drawbacks of implementing MBSE, how the U.S. can benefit from MBSE methodologies in nuclear energy, and the overall impacts of implementing MBSE in American nuclear energy operations and development. This paper seeks to highlight how the benefits of utilizing MBSE in nuclear energy operations outweigh the negative outcomes of doing so.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113752"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167574","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 design and analysis of the accident-tolerant fuel (ATF) application to VVER-1000 nuclear reactor as well as evaluation of dynamic parameters","authors":"S. Nasiri,&nbsp;G.R. Ansarifar","doi":"10.1016/j.nucengdes.2024.113814","DOIUrl":"10.1016/j.nucengdes.2024.113814","url":null,"abstract":"<div><div>In order to improve the safety of commercial nuclear reactors, this paper presents neutronic design and analysis of the accident-tolerant fuels (ATF) application to VVER-1000 nuclear reactor. The study commences by simulating a fuel assembly containing 2.4% enriched uranium.</div><div>The simulation of a VVER-1000 reactor core loaded with standard uranium dioxide (UO₂) fuel was then conducted using the MCNPX 2.6 code. This simulation yielded various neutronic and dynamic parameters, including the multiplication factor, excess reactivity, delayed neutron fraction, fuel and coolant temperature reactivity feedbacks, power peaking factor, and fuel burn-up. These values were subsequently compared with existing reference data.</div><div>Finally, the reactor core was simulated using accident-tolerant fuel. The resulting data was analyzed to identify the optimal combination of fuel and cladding. Throughout the study, the MCNPX 2.6 software was employed for neutronic core analysis. Comparative analysis revealed that uranium carbide (UC) offers superior safety margins in both neutronic and dynamic aspects, outperforming uranium dioxide (UO₂) and uranium mononitride (UN).</div><div>Simulations exploring the use of silicon carbide (SiC) and FeCrAl as alternative cladding materials within the reactor core demonstrated potential advantages over traditional Zirconium (Zr). Results indicate that application of ATF can improve cycle length, temperature reactivity feedbacks and power peaking factor compare to the reference core (UO₂ + Zr), significantly.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113814"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167577","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}