Nuclear Engineering and Design最新文献

{"title":"Explore the potential advantages of replacing UO2 with different thorium-based fuels in U.S. SCLWR","authors":"Nassar Alnassar ,&nbsp;Muneerah A. Al-Aqeel ,&nbsp;Sitah Alanazi ,&nbsp;Maha Algarawi ,&nbsp;Mohamed Y.M. Mohsen ,&nbsp;A. Abdelghafar Galahom","doi":"10.1016/j.nucengdes.2024.113751","DOIUrl":"10.1016/j.nucengdes.2024.113751","url":null,"abstract":"<div><div>This work searches for enhancing the efficiency of the supercritical light water reactor SCLWR by introducing different fuels based on thorium as an alternative fuel to conventional fuel. United States fuel assembly model of SCLWR has been modeled using Monte Carlo code MCNPX, ed. 2.7. The burnup results, k_inf, fuel composition and most importantly actinides and non-actinides have been examined and compared with that of the UO₂. The neutronic safety coefficients such as reactivity temperature coefficient and control rod worth have been calculated for the suggested fuel cycles. Also, the radial and axial neutron flux as well as power distribution have been analyzed to verify the validity of the proposed fuels. Assessment results demonstrate the viability of the proposed fuel. Reactivity measurements indicate that the introduced fuels have an economic benefit as using ²³²Th with ²³³U as a fuel increased the fuel cycle by 140 EFPD compared to the standard fuel cycle of UO₂. The reactivity temperature coefficients show that the proposed fuel provides a suitable level of passive safety as the fuel temperature coefficient (FTC) and the moderator temperature coefficient (MTC) values of the investigated fuel range from −1.32 to −3.34 pcm/K and −8.8 to −25.9 pcm/K, respectively. The power distribution indicates that there will be no melting of the fuel. This is because the power is not accumulated in specific fuel rods for the investigated fuels. The analysis of actinides concentrations shows that SCLWR is a good consumer of rgPu where it consumed more than 69% of the rgPu used in the fuel at the beginning of the fuel cycle.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113751"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167566","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":"Rhenium volatilization in SiO2-KNO3-KReO4 mixtures at elevated temperatures","authors":"Chenchen Niu ,&nbsp;Aiqing Chen ,&nbsp;Kaiwen Hu ,&nbsp;Kai Xu","doi":"10.1016/j.nucengdes.2024.113758","DOIUrl":"10.1016/j.nucengdes.2024.113758","url":null,"abstract":"<div><div>Due to its long half-life, high fission yield, and high mobility in the groundwater, Technetium-99 (Tc-99) is a troublesome nuclide in nuclear waste. Moreover, the primary concern is the high volatility of Tc during nuclear waste vitrification. Previous studies demonstrated that nitrate significantly affected the Tc volatilization during the vitrification of Tc-containing waste. However, identifying how nitrate affects Tc volatilization is challenging due to the complexity of the composition of nuclear waste. This study used a simplified SiO₂-KNO₃-KReO₄ feed with rhenium (Re) as a Tc non-radioactive surrogate to study the nitrate effect on the Re volatilization at elevated temperatures. The results showed that KNO₃ could facilitate Re volatilization at temperatures below 800 °C when the molar percentage of KNO₃ in the KNO₃-KReO₄ binary system exceeded 60 mol%, which was attributed to the eutectic reaction between KReO₄ and KNO₃ at 310 °C. However, the increased KNO₃ content could inhibit the Re volatilization at temperature above 1200 °C since adding KNO₃ could increase the amorphous phase content, where Re was encapsulated and hardly evaporated.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113758"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167573","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":"Editorial: 26th International Conference on Structural Mechanics in Reactor Technology","authors":"","doi":"10.1016/j.nucengdes.2024.113790","DOIUrl":"10.1016/j.nucengdes.2024.113790","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113790"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167576","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":"Comparison of neutronics performance of various TRISO fuels","authors":"Yan-Xin Chen ,&nbsp;Shin-Rong Wu ,&nbsp;Jason Chao ,&nbsp;Der-Sheng Chao ,&nbsp;Jhao-Yang Hong ,&nbsp;Jenq-Horng Liang","doi":"10.1016/j.nucengdes.2025.113825","DOIUrl":"10.1016/j.nucengdes.2025.113825","url":null,"abstract":"<div><div>TRISO (Tristructural Isotropic) particle fuel is a type of advanced nuclear fuel developed for High-Temperature Gas-Cooled Reactors (HTGRs). Its robust containment properties make TRISO particle fuel a promising candidate for Accident-Tolerant Fuels (ATFs) for next-generation reactors. In this study, the physics model of HTTR was established with the Monte Carlo code MCNP6.2 to evaluate the neutronic properties when utilizing various TRISO form fuels in High-Temperature Test Reactor (HTTR). The fuels considered in this study include UO₂, UC, and UN embedded in the core of TRISO particles. Moreover, the UC fuel is coated with TiN (UC_TiN), while the UN fuel is coated with ZrC (UN_ZrC). These two fuel types have been recognized as having manufacturing prospects, so ascertaining the effects of these coating materials on their neutronic performance before conducting practical applications is necessary. The results indicate that the isothermal temperature coefficients of the UC and UN fuels remained negative during the experimental operation, particularly −0.022 to −0.032 lower than the UO₂_TRISO fuel in the operational temperature, −0.016 to −0.043 lower in the condition when it exceeded the temperature of operational condition, shows the safety aspect of the utilization of ATFs. The effective multiplication factor of each fuel remained nearly constant during 660 EFPD; each fuel, accounting for the initial drop, has a <span><math><mrow><mi>Δ</mi><mi>k</mi><mo>/</mo><msub><mi>k</mi><mrow><mi>final</mi></mrow></msub><mo>&lt;</mo><mn>10</mn><mo>%</mo></mrow></math></span> indicating that the fuel replacement that occurred during the operation procedure maintained stability. Moreover, the neutron spectrum at the Beginning of Cycle (BOC) and End of Cycle (EOC) showed that the significant difference in the effective multiplication factor of UC_TRISO and UC_TiN was caused by the coating layers. The moderator-to-fuel ratio provides evidence that the difference between the fuels was caused by the fuel design instead of operational conditions, indicating that the heavy metal coating might not be suitable for HTGRs compared to TRISO coating. The spent fuel analysis is discussed to clarify fuel variations, which shows that the minor actinides of each fuel have increased about 4.2 %–27.7 % in comparison with the UO₂_TRISO, based on these neutronic mechanisms and calculation analysis, thus providing compelling evidence supporting the feasibility of applying ATFs in HTGRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113825"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167944","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":"Seismic finite element analysis of passive rectangular heat pipe heat exchange tank in nuclear power plant","authors":"Jiadong He ,&nbsp;Song Lu ,&nbsp;Jian Kang ,&nbsp;Tao Lu","doi":"10.1016/j.nucengdes.2024.113809","DOIUrl":"10.1016/j.nucengdes.2024.113809","url":null,"abstract":"<div><div>In this paper, an efficient and simple three-dimensional finite element seismic analysis method for rectangular ventilation heat exchange tank of nuclear power plant is proposed. Firstly, the liquid sloshing effect is simulated by the layered mass-spring model, and the liquid sloshing effect of three diameter methods under the simplified model of Housner and IITK-GSDMA is compared and analyzed. Secondly, in order to improve the analysis efficiency and maintain the key dynamic characteristics of the model, the three-dimensional calculation model is reasonably simplified by using the technology of intermediate surface extraction and beam extraction, and a fine three-dimensional finite element model is constructed which can capture the subtle stress changes under the action of seismic loads. Finally, the natural frequency of the rectangular tank is determined by the finite element modal analysis method, and the stress and strain of the rectangular tank under the actual seismic load is predicted by the response spectrum analysis method. The results show that the first-order shaking frequency calculated by the equal area diameter method based on Housner simplified model is consistent with the first-order shaking frequency simulated by the finite element method, which verifies the reliability and convenience of using the equal-area method to simplify the finite element seismic analysis of the rectangular water tank. This method can accurately and quickly predict the dynamic response of liquid under earthquake action, and provides a more scientific and reliable analysis means for seismic design and risk assessment of rectangular water tank, effectively improves the safety and reliability of rectangular water tank, and provides a strong support for the seismic design of key facilities such as nuclear power plants.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113809"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168322","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 evaluation of a coupled CFX-RELAP5 tool adopting experimental data from the TALL-3D facility","authors":"P. Cioli Puviani ,&nbsp;T. Del Moro ,&nbsp;B. Gonfiotti ,&nbsp;D. Martelli ,&nbsp;C. Ciurluini ,&nbsp;F. Giannetti ,&nbsp;R. Zanino ,&nbsp;M. Tarantino","doi":"10.1016/j.nucengdes.2024.113818","DOIUrl":"10.1016/j.nucengdes.2024.113818","url":null,"abstract":"<div><div>The interest for multiscale simulations for the thermal–hydraulic analysis of nuclear systems has increased with the grow of computational power and the need of complex three-dimensional analysis for the generation IV nuclear reactors. To perform analyses at different scales, coupling codes are expected to be a valuable solution. Particularly, the coupling between System Thermal Hydraulic (STH) and Computational Fluid Dynamics (CFD) codes promises to achieve the goal of a detailed local solution at CFD level, while keeping the speed of a STH approach elsewhere in the system, where 1D phenomena are predominant. The validation of new developed coupled tools requires specific experimental facilities in which three-dimensional phenomena affect the behaviour of the entire system. TALL-3D is a liquid Lead Bismuth Eutectic loop designed in such a way to produce experimental data for coupled STH and CFD codes validation. In this work, a Ansys CFX − RELAP5/Mod3.3 coupled tool is adopted to replicate a forced to natural circulation transition on the TALL-3D facility, comparing different coupling strategies regarding the spatial domain discretization (decomposition and overlapping) and time advancing scheme (explicit and semi-implicit). Results obtained with the coupled models show a clear advantage with respect to the RELAP5 standalone one. The analysis of the phenomena involved show the inability of the STH code in reproducing them, while the coupled tool proved to be a reliable solution thanks to its capability to take advantage of both the component and system scales analysis. Moreover, the impact of the selected coupling strategy on the stability of the tool is assessed, and the required computational time for the analysed transient is evaluated.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113818"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168370","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":"Next-generation solutions for water sustainability in nuclear power plants: Innovations and challenges","authors":"Ravikumar Jayabal","doi":"10.1016/j.nucengdes.2024.113757","DOIUrl":"10.1016/j.nucengdes.2024.113757","url":null,"abstract":"<div><div>Nuclear power plants (NPPs) are crucial for meeting global energy demands but face significant challenges due to their high water consumption, especially in water-scarce regions. These plants primarily use water for cooling, which can lead to substantial withdrawals and thermal pollution, impacting aquatic ecosystems. This article reviews advanced cooling technologies and wastewater treatment methods that aim to enhance water efficiency and reduce environmental impacts. Closed-loop systems and innovative water reclamation technologies, such as membrane filtration and reverse osmosis, are highlighted for their potential to minimize water withdrawal and improve reuse efficiency. Integrating artificial intelligence (AI) in hybrid cooling systems and using nanomaterials for water treatment are also discussed as promising solutions to enhance sustainability in NPPs. The paper emphasizes the importance of regulatory frameworks and economic considerations in implementing these technologies, aiming to ensure the long-term viability of nuclear energy as a low-carbon power source. The nuclear sector can contribute to sustainable water management and meet stricter environmental regulations by addressing these challenges.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113757"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168372","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":"Study on the design of unattended SCRS full-condition adaptive bypass flow systems","authors":"Bowen Zhang,&nbsp;Yizhuo Li,&nbsp;Haixu Zhu,&nbsp;Yunze Xue,&nbsp;Yuandong Zhang","doi":"10.1016/j.nucengdes.2024.113768","DOIUrl":"10.1016/j.nucengdes.2024.113768","url":null,"abstract":"<div><div>The Supercritical CO₂ Reactor System (SCRS), a member of the fourth-generation reactor family, offers promising prospects for practical application. The use of S-CO₂ as a working fluid leads to a more compact reactor design and higher energy conversion efficiency, positioning it as a strong contender for next-generation unattended intelligent reactor systems. However, the thermophysical properties of S-CO₂ exhibit dramatic changes near the critical point, placing higher demands on the control system to ensure SCRS stability. To address the challenge that PID control theory struggles to meet the complex control demands of supercritical reactors, this paper proposes an intelligent control-based approach. It utilizes neural networks to adaptively adjust the outputs of fuzzy control and PID control in the bypass flow control system. The optimized controller outperforms the original PI controller during normal operational transients and exhibits robust performance under accident conditions. The work presented in this paper provides valuable insights for the design and optimization of control systems in supercritical reactors for practical engineering applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113768"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169076","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":"Application of genetic algorithms to optimize fuel loading pattern and use of burnable absorbers to minimize power peaking and control excess reactivity in gas cooled fast reactor","authors":"Shohanul Islam","doi":"10.1016/j.nucengdes.2024.113802","DOIUrl":"10.1016/j.nucengdes.2024.113802","url":null,"abstract":"<div><div>This study aims to optimize the fuel loading pattern in gas-cooled fast reactors (GFRs) using genetic algorithms (GA) to reduce the power peaking factor, and in the GA-optimized model, several burnable absorbers (BA), namely Gd₂O₃, Er₂O₃, AmO₂, ZrB₂, HfO₂, Ta₂O₅, Dy₂O₃, Eu₂O₃, and Lu₂O₃ were deployed to control excess reactivity and further enhance power distribution. The GA-optimized fuel loading pattern significantly lowered the power peaking factor without compromising the effective multiplication factor. All GA + BA models, except for the Am and ZrB₂ models, successfully minimized both power peaking and excess reactivity. The Eu and Lu models demonstrated best result in reducing these parameters, though the Eu model’s shorter cycle length and larger reactivity swing make it less favorable. Gd, Er, Hf, Ta, and Dy models also displayed satisfactory performance in lowering both PPF and excess reactivity, and maintained criticality for over 2000 days with moderate reactivity swings. Neutronics analysis revealed that all models achieved harder neutron spectra and maintained uniform radial neutron flux distribution. Additionally, all models achieved satisfactory values for the effective delayed neutron fraction and fuel temperature coefficient, except Ta and Eu model. Overall, the Lu model emerged as the most favorable burnable absorber, as it achieved lower power peaking factor, reduced excess reactivity, good cycle length, moderate reactivity swing, satisfactory neutronics performance, and favorable beta effective and Doppler constant values.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113802"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169080","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 multi-fidelity multi-scale methodology to accelerate development of fuel performance codes","authors":"D. Pizzocri ,&nbsp;G. Zullo ,&nbsp;G. Petrosillo ,&nbsp;L. Luzzi ,&nbsp;F. Feria ,&nbsp;L.E. Herranz","doi":"10.1016/j.nucengdes.2024.113741","DOIUrl":"10.1016/j.nucengdes.2024.113741","url":null,"abstract":"<div><div>Multi-scale methodologies have been developed and applied successfully in the frame of nuclear fuel performance analyses, but the complexity of the tools involved hinders their extensive application. Gaps in modelling capabilities of specific input/outputs in particular limits code-to-code communication. In this work, we propose a multi-fidelity methodology to tackle this issue. The application presented here concerns the inclusion of a meso-scale module describing fission gas behaviour (SCIANTIX) in a fuel performance code (FRAPCON). A critical input parameter of the meso-scale module, the local hydro-static stress in the fuel, is not predicted by such fuel performance code, hence limiting this coupling. This gap is filled by using a second fuel performance code (TRANSURANUS) to construct a virtual dataset of local hydro-static stress values, on which an artificial neural network is trained and included in the FRAPCON/SCIANTIX coupled suite. This multi-fidelity methodology is demonstrated by simulating the Risø AN3 irradiation experiment.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113741"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169084","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}