Nuclear Engineering and Design最新文献

{"title":"Zircaloy-4 fuel pin failure under simulated loss-of-coolant-accident conditions: Creep and rupture","authors":"","doi":"10.1016/j.nucengdes.2024.113507","DOIUrl":"10.1016/j.nucengdes.2024.113507","url":null,"abstract":"<div><p>This paper presents a detailed investigation of the creep and rupture behavior of Zircaloy-4 (Zry-4) fuel claddings used in Indian Pressurized Heavy Water Reactors (IPHWR) under simulated Loss-Of-Coolant Accident (LOCA) conditions. Fuel claddings are pre-oxidized in the steam environment at 500 °C to mimic the oxygen pickup during normal reactor operation. The burst tests are then performed on these preoxidized tubes at different heating rates (55–115 K/s) and internal overpressures (15–45 bar) in the steam environment, creating LOCA-like scenario in a high burnup condition, wherein the claddings further oxidize while undergoing deformation and rupture. The burst stress correlation is developed for IPHWR claddings from the obtained burst temperature and oxygen concentration data. A burst criterion model is developed by solving available creep rate, oxidation rate, and phase transformation equations simultaneously to study the effect of various parameters on burst characteristics of the fuel cladding. The proposed burst criterion model agrees well with the present and previous experimental burst data. Also, the ballooning progression predicted from the model is validated with the present and previous experimental data. In addition, the effect of available correlations for the creep rate, phase boundary temperature, and oxidation factor on the burst characteristics has been presented.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141868109","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":"Preface for the ”OUTCOMES AND ACHIEVEMENTS FROM RESEARCHES ORIENTING THE FUTURE IN NUCLEAR FISSION TECHNOLOGY: Asia South-East (Vietnam, Thailand, Indonesia, Malaysia, Singapore, Bangladesh) and Australia”","authors":"","doi":"10.1016/j.nucengdes.2024.113500","DOIUrl":"10.1016/j.nucengdes.2024.113500","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141868108","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":"BORA4-PTS: Experimental reproduction of a Pressurized Thermal Shock, and building of numerical simulation with the CATHARE Code","authors":"","doi":"10.1016/j.nucengdes.2024.113497","DOIUrl":"10.1016/j.nucengdes.2024.113497","url":null,"abstract":"<div><p>When a break occurs in a nuclear reactor, a fast cooldown has to be down to prevent the melting of the core. This is done by the injection of cold water at 7 °C, in a pressurized vessel at 295 °C. This is a Pressurized Thermal Shock. To improve the safety of the nuclear reactor, an experimental facility was built to analyze the mix of hot and cold water in the downcomer of the vessel. This is the BORA4-PTS experiment. Salt water at 45 °C is injected into stagnant pure water at 20 °C, to represent the injection of cold water in hot water. Through this experiment, a scaling was establish to compare it with the reactor case. We then made a numerical simulation of this experimental facility with the CATHARE code. With this simulation come another scaling, in order to properly compare the numerical and the experimental results. The numerical simulations give results that are very similar to the experimental ones. With this experiment, we show the excellent capacity of CATHARE to simulate and model the complex thermohydraulic inside a downcomer.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0029549324005971/pdfft?md5=524c6451272a42ca4fad2a626c6251ef&pid=1-s2.0-S0029549324005971-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141867993","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":"Validation of CFD RANS of an internally heated natural convection in a hemispherical geometry","authors":"","doi":"10.1016/j.nucengdes.2024.113471","DOIUrl":"10.1016/j.nucengdes.2024.113471","url":null,"abstract":"<div><p>In the context of severe accidents, one mitigation strategy that has been shown to work for low-to-intermediate power reactors is the In-Vessel Melt Retention (IVMR) of molten corium. For this reason, several efforts have been put forward to make this strategy feasible for high power reactors. In particular, the aim of the European H2020 IVMR project was to evaluate and improve current modeling strategies, such as the use of Computational Fluid Dynamics (CFD) codes for the prediction of flow and heat transfer in a homogeneous corium pool. Due to evident limitations, the validation was mainly performed against an available water-based experimental data, rather than a corium mixture. In order to overcome this limitation, complementary high fidelity numerical simulations, in the form of Direct Numerical Simulation (DNS), have been performed recently and are used in the current work as a reference for the validation purposes of the Reynolds Averaged Navier–Stokes (RANS) approach. More specifically, RANS numerical simulations of a three-dimensional hemispherical configuration are performed using the STAR-CCM+ software. Consistent with the DNS approach, the Boussinesq assumption is used to characterize the internally heated (IH) natural convection problem. The flow conditions correspond to a Rayleigh number of <span><math><mrow><mn>1</mn><mo>.</mo><mn>6</mn><mi>⋅</mi><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>11</mn></mrow></msup></mrow></math></span> and a Prandtl number of 0.5. Several turbulence models available in STAR-CCM+, which are generally used for buoyancy driven flows, are compared and evaluated against the DNS results, in terms of velocity, temperature, buoyancy production of the turbulent kinetic energy and heat flux. Reasonable results are obtained by the RANS models, especially in predicting the main qualitative features of the flow configuration, such as thermal stratification, fast descending flow on the curved walls and high turbulence at the top of the domain. The main divergence between RANS and DNS is observed in the bulk region, where all the RANS computations present strong recirculation, while an extended nearly stagnant zone is predicted by DNS calculations. A quantitative analysis is performed as well, highlighting the limitations of the RANS approaches, especially for the turbulent heat flux modeling, and the need for the development of more advanced models as potential future efforts.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141867972","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 review of catalyst material for hydrogen mitigation systems in nuclear facilities","authors":"","doi":"10.1016/j.nucengdes.2024.113481","DOIUrl":"10.1016/j.nucengdes.2024.113481","url":null,"abstract":"<div><p>The Fukushima incident starkly underscores the imperative need to address the substantial safety risks posed by hydrogen explosions in various industrial systems utilizing hydrogen. An explosion risk arises when the concentration of hydrogen in a mixed gas surpasses 4%. Catalytic hydrogen combustion, characterized by its enhanced efficiency and safety, has emerged as a potent strategy to alleviate the detrimental effects of hydrogen explosions. This paper offers an exhaustive review of catalyst material for catalytic hydrogen combustion, encompassing diverse catalysts, and delineates the current research trajectory concerning catalyst design, fabrication, and development methodologies in a systematic manner. This review encapsulates the deleterious impacts of toxic substances—such as water vapor, carbon monoxide, iodine compounds, and fire combustion products—that may be present in nuclear facilities on catalysts and the implications of isotopic effects that warrant particular scrutiny in these settings. Finally, potential avenues for future research are suggested to alleviate hydrogen hazards in nuclear plants through the use of CHC.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141868102","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":"Preface for special issue “Outcomes and achievements from researches orienting the future in nuclear fission technology: Russia, Kazakhstan, Uzbekistan”","authors":"","doi":"10.1016/j.nucengdes.2024.113495","DOIUrl":"10.1016/j.nucengdes.2024.113495","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141873110","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":"Foreword for Special Issue NFT-02: Outcomes and achievements from researches orienting the future of Nuclear Fission Technology in Belgium","authors":"","doi":"10.1016/j.nucengdes.2024.113491","DOIUrl":"10.1016/j.nucengdes.2024.113491","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141873111","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 Seismic Probabilistic Risk Assessment of Nuclear Power Plants Using Advanced Structural Methodologies","authors":"","doi":"10.1016/j.nucengdes.2024.113416","DOIUrl":"10.1016/j.nucengdes.2024.113416","url":null,"abstract":"<div><p>The conditional core damage probability (CCDP) of a representative nuclear power plant is estimated for a beyond design basis earthquake (BDBE) using state-of-the-art structural models within a dynamic probabilistic risk assessment (DPRA) framework. Randomness of seismic excitation and uncertainty of structural parameters are considered using a simulation-based approach. Finite element models are developed for the structure and a liquid container (hydro-accumulator), and fluid–structure interaction is considered with the arbitrary Lagrangian-Eulerian method. The CCDP is evaluated through a time-dependent event tree. Also, correlation among multiple hydro-accumulators is investigated. The results show that operator action and implementation time of mobile safety equipment are critical under BDBE in case of loss of offsite power following an earthquake. It was also found necessary to adopt a DPRA approach to determine the available time for action and core damage timing probabilistically.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141868098","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":"Performance analysis of the temperature-upgraded flash-driven low-temperature advanced natural circulation heating reactor system","authors":"","doi":"10.1016/j.nucengdes.2024.113503","DOIUrl":"10.1016/j.nucengdes.2024.113503","url":null,"abstract":"<div><p>To address the current deficiencies in outlet temperature and thermal power of low-temperature heating reactors while ensuring safety and economic viability, this study introduces the Temperature-Upgraded Flash-driven Low-temperature Advanced Natural Circulation Heating Reactor (TU-FLANC). The FLANC system innovatively utilizes the flashing phenomenon of the coolant in the rising channel to significantly increase the coolant circulation flow rate and thus enhance thermal power at atmospheric pressure. The TU system employs an Absorption Heat Pump (AHP) to upgrade the temperature of the reactor’s output heat. The two systems are interconnected via a Coupled System Heat Exchanger (CSHEX), achieving an upgrade in reactor thermal power and outlet temperature at atmospheric pressure. To evaluate the performance of the TU-FLANC system, a mathematical model of the system was established, and a computational program was developed. The impact of key parameters such as evaporator temperature, condenser temperature, and solution concentration on system performance was analyzed. The results indicate that the evaporator temperature and solution concentration have the most significant impact on the system’s coefficient of performance (COP) and the coefficient of performance considering pump work (COPW). Through Differential Evolution (DE) algorithm optimization, the optimal solution concentration combinations were determined to maximize COP and COPW under different temperature upgrade demands. For a temperature upgrade demand of 50 °C, the optimal solution concentration combinations are 40.02 % and 57.48 %, with corresponding COP and COPW values of 0.5282 and 0.4886, respectively. The research findings highlight the significant innovative potential of the TU-FLANC system in enhancing heat power and outlet temperature.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141868097","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":"Phase-field simulation of fission bubbles formation in composite ceramic nuclear fuel","authors":"","doi":"10.1016/j.nucengdes.2024.113485","DOIUrl":"10.1016/j.nucengdes.2024.113485","url":null,"abstract":"<div><p>A phase-field model was developed to simulate the evolution of fission bubbles in UN-UO₂ composite ceramic nuclear fuel. The model was firstly used to simulate the formation of UO₂ intracrystalline bubbles, and the change of porosity obtained from the simulation was compared with the experimental data. The bubble evolution of UN-UO₂ composite fuels in the UN and UO₂ phase regions was then investigated for different burnups and doping concentrations. It was found that the nucleation rate of bubbles in the UN phase was slower than that in the UO₂ phase, but the growth rate of bubbles in the UN phase was faster, so that the bubbles in the UO₂ phase had a larger number density at low fuel consumption. It is specifically analyzed how the surface energy, diffusion coefficient and uranium density affect the difference in bubble evolution between the two phases.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-07-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141868099","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}