Annals of Nuclear Energy最新文献_第10页

Loss-of-heat-sink transient simulation with RELAP5/Mod3.3 code for the ATHENA facility 使用 RELAP5/Mod3.3 代码对 ATHENA 设施进行散热损失瞬态模拟

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-25 DOI: 10.1016/j.anucene.2024.110948

T. Del Moro , F. Giannetti , P. Cioli Puviani , I. Di Piazza , D. Diamanti , M. Tarantino

{"title":"Loss-of-heat-sink transient simulation with RELAP5/Mod3.3 code for the ATHENA facility","authors":"T. Del Moro , F. Giannetti , P. Cioli Puviani , I. Di Piazza , D. Diamanti , M. Tarantino","doi":"10.1016/j.anucene.2024.110948","DOIUrl":"10.1016/j.anucene.2024.110948","url":null,"abstract":"<div><div>ATHENA (Advanced Thermal-Hydraulic Experiment for Nuclear Applications) is a large multipurpose pool-type lead-cooled facility under construction at the Mioveni site in Romania. It has been identified by the FALCON (Fostering ALfred CONstruction) Consortium to characterize large to full-scale ALFRED components, to conduct integral tests, and to investigate the main thermal–hydraulic phenomena inherent in pool-type systems. ATHENA is representative of ALFRED in terms of the difference in height of the thermal barycenters of the heat source and heat sink, i.e., 3.3 m, in order to reproduce the buoyancy forces in the system. Similar to ALFRED’s design, ATHENA minimizes thermal stratification within the main vessel even under natural circulation conditions, through an internal structure referred to as “barrel”. This structure directs the fluid flow towards the main vessel, preventing fluid stagnation near the vessel itself. The paper initially provides a steady-state thermal–hydraulic characterization of the facility, including details of the numerical model developed using the RELAP5/Mod3.3 thermal–hydraulic code. Then, focus is given to the transient analysis considering as a reference scenario a Loss-of-Heat-Sink (LOHS) accidental transient. In this scenario, the Main Circulation Pump (MCP) is assumed to remain operational while the Core Simulator (CS) is deactivated once the lead temperature at the Main Heat Exchanger (MHX) outlet reaches a predefined threshold. A sensitivity analysis is conducted with set points of 430 °C, 450 °C, 470 °C, and 490 °C, assessing the system’s response following MHX isolation from the secondary loop. The study evaluates the impact of different CS deactivation set points on reactor SCRAM delay (reducing CS power to a level representative of decay heat) as well as on system maximum and minimum temperatures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110948"},"PeriodicalIF":1.9,"publicationDate":"2024-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142319016","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}

引用次数: 0

Coupling of MELCOR with surrogate model for quench estimation of conical debris beds 将 MELCOR 与用于锥形碎片床淬火估算的代用模型相耦合

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-24 DOI: 10.1016/j.anucene.2024.110933

Wanhong Wang, Weimin Ma

{"title":"Coupling of MELCOR with surrogate model for quench estimation of conical debris beds","authors":"Wanhong Wang, Weimin Ma","doi":"10.1016/j.anucene.2024.110933","DOIUrl":"10.1016/j.anucene.2024.110933","url":null,"abstract":"<div><div>The MELCOR code as a severe accident simulation tool does not have the capability to capture the quench process of a debris bed which may form in the wet cavity during a severe accident of light water reactors (LWRs). Although the coupled MELCOR/COCOMO simulation could overcome the limitation (Chen et al., 2022), the calculation time was explosively escalated due to mechanistic modeling of debris bed thermal-hydraulics in COCOMO. To suppress the computational cost, a surrogate model (SM) was developed in our previous study (Wang et al., 2023), and its coupling with MELCOR could realize a quick estimation of the quench process of one-dimensional debris beds. The present study is an extension of the previous work, aiming at the development of a new surrogate model for the quench process of two-dimensional conical debris beds. The new surrogate model (SM) was based on artificial neural networks (ANNs) and trained by the database from COCOMO calculations of various conical debris beds quenched in the reactor cavity of a Nordic boiling water reactor (BWR). The MELCOR was then coupled with the new SM to simulate a postulated station blackout (SBO) scenario in the BWR. The results show that the coupled MELCOR/SM simulation could provide similar ex-vessel debris bed quench period and containment pressure/temperature trends as the coupled MELCOR/COCOMO. Compared with the MELCOR standalone calculation, the coupled calculations predicted earlier points of time for water pool saturation and containment venting, since the heat transfer from conical debris bed to water pool is faster in the coupled simulations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110933"},"PeriodicalIF":1.9,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0306454924005966/pdfft?md5=30a8d621b7bdd294c65c323d9f6699a8&pid=1-s2.0-S0306454924005966-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142311266","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}

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Experimental study on onset of nucleate boiling in wide-ranged parameters for narrow rectangular channels 窄矩形水道宽参数条件下核沸腾发生的实验研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-24 DOI: 10.1016/j.anucene.2024.110935

Bo Kuang, Yu Zhao, Gang Wang, Chengque Cao, Pengfei Liu

{"title":"Experimental study on onset of nucleate boiling in wide-ranged parameters for narrow rectangular channels","authors":"Bo Kuang, Yu Zhao, Gang Wang, Chengque Cao, Pengfei Liu","doi":"10.1016/j.anucene.2024.110935","DOIUrl":"10.1016/j.anucene.2024.110935","url":null,"abstract":"<div><div>The onset of nucleate boiling (ONB), which marks the emergence of nucleate boiling, is an important transition point in the boiling curve. For exploring the influence of geometric and thermodynamic parameters on ONB in rectangular narrow channels, a detailed experimental study is conducted to investigate ONB under wide range of parameters. The experimental parameters range is pressure of 0.1–5.5 MPa, mass flux of 200–2000 kg/m<sup>2</sup>s, inlet subcooling of 10–150 K. According to the experimental results, the location of ONB is identified based on the axial distribution of wall temperature, and the influence of various parameters on ONB in narrow rectangular channels is analyzed. It is found that heat flux, pressure, mass flux, and the gap size of the channel have a significant impact on ONB. By comparing the computed results of existing correlations, it is evident that there is a deviation, which can be attributed to the narrow range of experimental parameters in previous studies. Finally, a new ONB model is developed based on basic equations proposed by Hsu and the distribution of liquid temperature, taking into account the influence of mass flux and the enhanced heat transfer results from surrounding bubbles to correct the liquid temperature. The new correlation accurately describes the impact of each parameter and is in good agreement with the current experimental results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110935"},"PeriodicalIF":1.9,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142315861","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}

引用次数: 0

Improvement of Geant4 Neutron-HP package: Unresolved resonance region description with probability tables 改进 Geant4 中子 HP 软件包：用概率表描述未解决的共振区域

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-24 DOI: 10.1016/j.anucene.2024.110914

M. Zmeškal , L. Thulliez , P. Tamagno , E. Dumonteil

{"title":"Improvement of Geant4 Neutron-HP package: Unresolved resonance region description with probability tables","authors":"M. Zmeškal , L. Thulliez , P. Tamagno , E. Dumonteil","doi":"10.1016/j.anucene.2024.110914","DOIUrl":"10.1016/j.anucene.2024.110914","url":null,"abstract":"<div><div>Whether for shielding applications or for criticality safety studies, solving the neutron transport equation with good accuracy requires to take into account the resonant structure of cross sections in part of the Unresolved Resonance Region (URR). In this energy range even if the resonances can no longer be resolved experimentally, neglecting them can lead to significant numerical biases, namely in flux-based quantities. In Geant4, low energy neutrons are transported using evaluated nuclear data libraries handled by the Neutron High-Precision (Neutron-HP) package. In the version 11.01.p02 of the code, the URR can only be described by average smooth cross sections that do not take into account the statistical resonant structure of the cross sections. To overcome this shortcoming, the treatment of the URR with the use of the probability table method has been implemented in Geant4 and successfully validated with the reference Monte Carlo neutron transport codes MCNP6 (version 6.2) and Tripoli-4® (version 12). These developments will be taken into account in the next release of Geant4. All the validations of Geant4 have been performed with probability tables generated from both the NJOY and CALENDF pre-processing tools. Therefore Geant4 now has this unique feature to study the relative impact of the strategies involved during the production of probability table by the two pre-processing codes. This has been used to show that self-shielding is important also for inelastic cross sections in the example of <sup>238</sup>U. The tool to generate probability tables usable by Geant4 either from NJOY or from CALENDF is made available on a dedicated GitLab repository and will be included in Geant4.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110914"},"PeriodicalIF":1.9,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142315862","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}

引用次数: 0

Improvement of the determination of radioactivity concentrations of short-lived radioisotopes with the time-dependent leakage model of VVER-type nuclear reactor 利用 VVER 型核反应堆随时间变化的泄漏模型改进短寿命放射性同位素放射性浓度的测定方法

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-23 DOI: 10.1016/j.anucene.2024.110929

Csongor Kristóf Szarvas , Gábor Radócz , Anita Gerényi , Péter Jakab Rozmanitz , Imre Szalóki

{"title":"Improvement of the determination of radioactivity concentrations of short-lived radioisotopes with the time-dependent leakage model of VVER-type nuclear reactor","authors":"Csongor Kristóf Szarvas , Gábor Radócz , Anita Gerényi , Péter Jakab Rozmanitz , Imre Szalóki","doi":"10.1016/j.anucene.2024.110929","DOIUrl":"10.1016/j.anucene.2024.110929","url":null,"abstract":"<div><div>The leak-free operation is a fundamental necessity for the entire duration of each nuclear reactor cycle. While the physical and chemical processes resulting from fuel failures typically do not avert the safe operation of the reactor, they can have direct operational and economic consequences. The state of leak-free operation in the reactor core is monitored on-line by gamma spectra measurements of the primary coolant. Estimating the burnup and leakage parameters of defective fuel elements is achieved through calculations employing relevant leakage models. This paper presents a novel approach to assessing the time-dependent sample delay time, which is used to calculate the time-dependent radioactivity concentrations of the fission products in the primary coolant from their number of nuclei calculated by the time-dependent leakage model. The performance of three models with different basic assumptions was individually and collectively investigated. The new sample delay time model allows a more accurate estimation of the time-dependent radioactivity concentration, especially of the short-lived <sup>137</sup>Xe, <sup>138</sup>Xe, <sup>138</sup>Cs, and <sup>134</sup>I isotopes in the primary coolant.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110929"},"PeriodicalIF":1.9,"publicationDate":"2024-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0306454924005929/pdfft?md5=aff1b1b231641baad91d94c3d7924725&pid=1-s2.0-S0306454924005929-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142311265","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}

引用次数: 0

Nuclide number density prediction in the lattice physics calculation based on Dynamic mode decomposition 基于动态模式分解的晶格物理计算中的核素数量密度预测

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-20 DOI: 10.1016/j.anucene.2024.110924

Shuai Qin , Qian Zhang , Yunfei Zhang , Pengchao Xue , Zhuo Li

{"title":"Nuclide number density prediction in the lattice physics calculation based on Dynamic mode decomposition","authors":"Shuai Qin , Qian Zhang , Yunfei Zhang , Pengchao Xue , Zhuo Li","doi":"10.1016/j.anucene.2024.110924","DOIUrl":"10.1016/j.anucene.2024.110924","url":null,"abstract":"<div><p>Burnup analysis in nuclear reactors requires iterative computation of neutron transport and fuel depletion, which is computationally intensive, particularly for large-scale scenarios. This study introduces an innovative approach leveraging the Dynamic Mode Decomposition (DMD) algorithm to predict the temporal evolution of nuclide densities. By identifying and utilizing the DMD modes and eigenvalues from snapshots of nuclide density, this method aims to alleviate the computational demands of the coupled transport and burnup calculations. Firstly, the methodology selects the key reactivity-contributing nuclides to evaluate the correlation between the complexity of the reduced-order model and the precision of predictions. Subsequently, an optimized reduced-order model is employed for forecasting nuclide densities in a pin-cell. In most cases, DMD predicts more accurately than traditional quadratic extrapolation methods. Moreover, the DMD algorithm demonstrates commendable accuracy in predicting the nuclide density distribution within a PWR fuel assembly, suggesting its promising potential for reactor burnup analysis applications.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110924"},"PeriodicalIF":1.9,"publicationDate":"2024-09-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142274386","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}

引用次数: 0

Excitation functions for fast neutron induced reactions on Al, Zr, In and Au 铝、锆、铟和金上快中子诱导反应的激发函数

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-20 DOI: 10.1016/j.anucene.2024.110925

B. Liu , G. Tian , R. Han , F. Shi , H. Sun , Z. Chen , Z. Zhang , Q. Li , P. Luo

{"title":"Excitation functions for fast neutron induced reactions on Al, Zr, In and Au","authors":"B. Liu , G. Tian , R. Han , F. Shi , H. Sun , Z. Chen , Z. Zhang , Q. Li , P. Luo","doi":"10.1016/j.anucene.2024.110925","DOIUrl":"10.1016/j.anucene.2024.110925","url":null,"abstract":"<div><p>Cross sections of the <sup>27</sup>Al(n,<span><math><mi>α</mi></math></span>)<sup>24</sup>Na, <sup>96</sup>Zr(n,2n)<sup>95</sup>Zr, <sup>115</sup>In(n,p)<sup>115<em>g</em></sup>Cd, <sup>115</sup>In(n, 2n)<sup>114<em>m</em></sup>In and <sup>197</sup>Au(n,2n)<sup>196/196m2</sup>Au reactions induced by D-T neutrons are presented with the activation method and off-line <span><math><mi>γ</mi></math></span>-ray spectrometry technique. Uncertainty propagation and correlation of the cross sections were estimated using covariance analysis. It shows that, our results are consistent with most of the previous literature data of EXFOR library. Experimental values, including previous literature data, are compared with evaluated nuclear data of the CENDL-3.2, BROND-3.1, ENDF/B-VIII.0, IRDFF-II, JENDL-5 and JEFF-3.3 libraries. Besides, these excitation functions were theoretically calculated by using the TALYS-2.0 code up to the neutron energy of 20 MeV. It shows that significant discrepancies were found between experiment data and those of calculated results and evaluated data.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110925"},"PeriodicalIF":1.9,"publicationDate":"2024-09-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142274385","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}

引用次数: 0

Phase-field simulation of recrystallization and calculation of the effective thermal conductivity of polycrystalline UO2 多晶二氧化铀再结晶的相场模拟和有效热导率计算

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-19 DOI: 10.1016/j.anucene.2024.110918

Yanbo Jiang , Wenlong Shen , Yongxiao La , Xun Lan , Wenbo Liu

{"title":"Phase-field simulation of recrystallization and calculation of the effective thermal conductivity of polycrystalline UO2","authors":"Yanbo Jiang , Wenlong Shen , Yongxiao La , Xun Lan , Wenbo Liu","doi":"10.1016/j.anucene.2024.110918","DOIUrl":"10.1016/j.anucene.2024.110918","url":null,"abstract":"<div><p>During the operational lifespan of uranium dioxide (UO<sub>2</sub>) fuel, the emergence of a specific process termed recrystallization may transpire. The influence of recrystallization on the thermal conductivity of the fuel holds paramount significance, bearing direct implications for both safety and economic considerations. In the current investigation, a phase-field model incorporating an explicit nucleation model for recrystallized grains was formulated to study the formation and growth of recrystallized grains within polycrystalline UO<sub>2</sub>. The simulations conducted in this study revealed that the kinetics of recrystallization adhered to the empirical equation, and the observed variation in grain size during recrystallization exhibited concordance with experimental data. To elucidate the variation in thermal conductivity during recrystallization, a thermal conductivity model based on the microstructure generated through phase-field simulations was employed. The relationship between grain boundary (GB) thermal resistance and phase-field simulation parameters has been determined through empirical formulas. The simulated values of thermal conductivity during recrystallization demonstrated a commendable agreement with empirical functions. By comparing the computational results of thermal conductivity with or without recrystallization, it is proven that recrystallization is beneficial to the effective thermal conductivity because the increase in thermal conductivity due to the elimination of defects by recrystallization exceeds the decrease in thermal conductivity due to the introduction of large area GBs.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110918"},"PeriodicalIF":1.9,"publicationDate":"2024-09-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142274384","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}

引用次数: 0

Steady-state analysis of Xi’an Pulse Reactor based on the multi-physics coupling method 基于多物理场耦合方法的西安脉冲反应堆稳态分析

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-18 DOI: 10.1016/j.anucene.2024.110922

Tianliang Hu , Duoyu Jiang , Xinyi Zhang , Zhaohao Wang , Lipeng Wang , Lu Cao , Da Li , Lixin Chen

引用次数: 0

Applicability domain and gaps of SNF decay heat validation data – A similarity-based approach 核燃料衰变热验证数据的适用领域和差距--基于相似性的方法

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-18 DOI: 10.1016/j.anucene.2024.110905

Ahmed Shama , Stefano Caruso , Dimitri Rochman

{"title":"Applicability domain and gaps of SNF decay heat validation data – A similarity-based approach","authors":"Ahmed Shama , Stefano Caruso , Dimitri Rochman","doi":"10.1016/j.anucene.2024.110905","DOIUrl":"10.1016/j.anucene.2024.110905","url":null,"abstract":"<div><p>Decay heat measurements on spent nuclear fuel (SNF) provide a basis for code validation. Their applicability domain (AD) and gaps, which are the focus of this study, are not commonly discussed in the literature. The analyzed validation data are based on measurements at the Clab facility and on calculations using the Polaris and ORIGEN codes of the SCALE code system. Bias-predicting machine learning (ML) models are applied: random forest and weighted k-nearest neighbors. The models weigh the similarity between the cases, expressed using correlations. The learning curves are studied by examining the prediction error versus the sample size and the similarity coefficient. The obtained error reduction at higher similarity coefficients supports the argument that the similarity or correlation is informative. However, a marginal error reduction is expected from increasing the validation data size from its current status. Following this, a validation AD is proposed as a range of SNF characteristics within which the validation data and the ML models are observed and tested. Within the AD, different levels of error, i.e., safety margins and conservatism, were evaluated. Beyond the AD, validation gaps exist. Examination of light-water reactor SNF applications indicates that the validation coverage is absent in both MOX fuel and short cooling, diminishes rapidly at higher burnup for low-enrichment fuel, and extends with burnup for high-enrichment cases. Additional measurements are justified to reduce conservatism or achieve validation coverage in applications. A case study of typical UO<sub>2</sub> and MOX SNF applications is analyzed. It is shown that a few tens of optimally selected measurements from both SNF types are necessary to complete validation coverage in numerous applications.</p></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110905"},"PeriodicalIF":1.9,"publicationDate":"2024-09-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0306454924005681/pdfft?md5=55d8e2f442073aa564f9095ccd7b7e04&pid=1-s2.0-S0306454924005681-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142241591","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}

引用次数: 0