Annals of Nuclear Energy最新文献

{"title":"Research on the high-performance computing method for the neutron diffusion spatiotemporal kinetics equation based on the convolutional neural network","authors":"Xiangyu Li,&nbsp;Heng Xie","doi":"10.1016/j.anucene.2024.110943","DOIUrl":"10.1016/j.anucene.2024.110943","url":null,"abstract":"<div><div>Due to the uncertainty of computational results and the lack of interpretability of models in solving physical field equations in current deep learning, this paper designs a convolutional neural network that can be used to solve the neutron diffusion spatiotemporal kinetics equation in polar and cylindrical coordinate systems. This algorithm directly utilizes the macroscopic cross-section of the material without using the lattice homogenization method, replaces the finite volume method with the extended matrices, and solves the extended matrices using the convolutional kernels instead of the iterative algorithms. Taking the simplified Tsinghua High Flux Reactor (THFR) as an example, the feasibility of the algorithm is verified on the PyTorch platform and compared with the calculation results of the source iteration method running on the GPU. The calculation results show that when the number of grids in the radial and axial sections of the simplified THFR model is 804,600 and 3,576,000, respectively, and the algorithm is iterated 3000 times, the normalized power of the convolutional neural network and the source iteration method converges to 10⁻¹⁰, and the maximum point by point error of the neutron flux density of the above two algorithms converges to 10⁻⁵. The computational time consumed by the convolutional neural network is approximately 880.64 s and 3729.62 s, which reduces the computational time by 4.66% and 5.05% compared to the GPU parallel accelerated source iteration method, and the former consumes 43.75% less memory compared to the latter. The convolutional neural network is mainly used as the virtual physics engine for the THFR digital twin system, in addition to solving the neutron diffusion spatiotemporal kinetics equation and further improving computational speed. The algorithm directly utilizes the neutron macroscopic cross-section of the material to calculate the neutron flux density distribution without using the lattice homogenization, providing theoretical guidance and algorithm support for developing the high-precision multi-physical field coupling model.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323276","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":"Pressurized water reactor fuel corrosion-related unidentified deposit and its related safety issues – II. Corrosion product deposition and heat transfer modeling","authors":"Yan Liu ,&nbsp;Jiajie Chen ,&nbsp;Guolian Wang ,&nbsp;Hui He ,&nbsp;Tengfei Zhang ,&nbsp;Xiaojing Liu","doi":"10.1016/j.anucene.2024.110932","DOIUrl":"10.1016/j.anucene.2024.110932","url":null,"abstract":"<div><div>CRUD depositions on fuel cladding are the main cause of power shift and localized corrosion in nuclear power plants. This paper is the second of a three-part study concerning the deposition of corrosion products and its related safety issues. In this paper, analytical modules are proposed to predict CRUD growth and internal heat and mass transfer. CRUD growth depends on dynamic balance between corrosion product deposition, flow erosion and chemical equilibrium. In the multi-module iteration, the CRUD thickness is updated first followed by internal temperature and concentration fields. Temperature affects the chemical equilibrium, deposition and erosion equilibrium on CRUD surfaces. The accuracy and reliability of the coupling method are verified by experimental results. The difference of effective thermal conductivity between previous experimental results and calculation results is less than 0.4384 W/(m × K) and the cladding temperature relative error between WALT Loop results and calculation results is less than 1 %. The influences of operation conditions are evaluated. Coolant with lower pH reduces corrosion product solubility leading to high CRUD thickness. The main source of CRUD growth is from soluble precipitation, because CRUD depositions formed from soluble precipitation are thicker than those from the insoluble particles of the same concentration. High heat flux increases CRUD growth, internal wick boiling and boron hideout. Hydrogen in reactor application range has a minimal meaningful effect on CRUD growth, wick boiling and boron hideout. This study provides a precise method for further understanding CRUD growth and its internal multi-physical phenomena to alleviate CRUD-related safety issues.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142319015","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":"Loss-of-heat-sink transient simulation with RELAP5/Mod3.3 code for the ATHENA facility","authors":"T. Del Moro ,&nbsp;F. Giannetti ,&nbsp;P. Cioli Puviani ,&nbsp;I. Di Piazza ,&nbsp;D. Diamanti ,&nbsp;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":null,"pages":null},"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}

{"title":"Coupling of MELCOR with surrogate model for quench estimation of conical debris beds","authors":"Wanhong Wang,&nbsp;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":null,"pages":null},"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}

{"title":"Experimental study on onset of nucleate boiling in wide-ranged parameters for narrow rectangular channels","authors":"Bo Kuang,&nbsp;Yu Zhao,&nbsp;Gang Wang,&nbsp;Chengque Cao,&nbsp;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/m²s, 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":null,"pages":null},"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}

{"title":"Improvement of Geant4 Neutron-HP package: Unresolved resonance region description with probability tables","authors":"M. Zmeškal ,&nbsp;L. Thulliez ,&nbsp;P. Tamagno ,&nbsp;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 ²³⁸U. 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":null,"pages":null},"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}

{"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 ,&nbsp;Gábor Radócz ,&nbsp;Anita Gerényi ,&nbsp;Péter Jakab Rozmanitz ,&nbsp;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 ¹³⁷Xe, ¹³⁸Xe, ¹³⁸Cs, and ¹³⁴I isotopes in the primary coolant.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"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}

{"title":"Nuclide number density prediction in the lattice physics calculation based on Dynamic mode decomposition","authors":"Shuai Qin ,&nbsp;Qian Zhang ,&nbsp;Yunfei Zhang ,&nbsp;Pengchao Xue ,&nbsp;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":null,"pages":null},"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}

{"title":"Excitation functions for fast neutron induced reactions on Al, Zr, In and Au","authors":"B. Liu ,&nbsp;G. Tian ,&nbsp;R. Han ,&nbsp;F. Shi ,&nbsp;H. Sun ,&nbsp;Z. Chen ,&nbsp;Z. Zhang ,&nbsp;Q. Li ,&nbsp;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 ²⁷Al(n,<span><math><mi>α</mi></math></span>)²⁴Na, ⁹⁶Zr(n,2n)⁹⁵Zr, ¹¹⁵In(n,p)^{115<em>g</em>}Cd, ¹¹⁵In(n, 2n)^{114<em>m</em>}In and ¹⁹⁷Au(n,2n)^196/196m2Au 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":null,"pages":null},"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}

{"title":"Phase-field simulation of recrystallization and calculation of the effective thermal conductivity of polycrystalline UO2","authors":"Yanbo Jiang ,&nbsp;Wenlong Shen ,&nbsp;Yongxiao La ,&nbsp;Xun Lan ,&nbsp;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 (UO₂) 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 UO₂. 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":null,"pages":null},"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}