Annals of Nuclear Energy最新文献

{"title":"On the multigroup albedo boundary conditions for discrete ordinates nuclear reactor global calculations in X,Y-rectangular geometry","authors":"C.E.A. Nunes ,&nbsp;H. Alves Filho ,&nbsp;F.C. Silva ,&nbsp;L.R.C. Moraes ,&nbsp;R.C. Barros","doi":"10.1016/j.anucene.2025.111540","DOIUrl":"10.1016/j.anucene.2025.111540","url":null,"abstract":"<div><div>Presented here is a study on the use of approximate albedo boundary conditions, for numerically solving multigroup discrete ordinates (S_N) neutron transport eigenvalue problems, in two-dimensional Cartesian geometry for nuclear reactor global calculations in homogenized assembly chessboard models. A matrix operator, referred to as multigroup albedo matrix, approximates the non-multiplying regions that typically surround the cores of nuclear reactors (e.g., baffle and reflector), by neglecting the transverse leakage terms that arise from transverse integrations of the X, Y-geometry multigroup S_N neutron transport equations within these regions. These approximate albedo matrices are obtained through convenient manipulations of the coarse-mesh Spectral Greeńs Function (SGF) method’s auxiliary equations and are applied to the traditional fine-mesh Diamond Difference (DD) method in reactor global calculations, aiming at shortening the computer running time by substituting the non-multiplying regions around the core through the offered albedo boundary conditions. Numerical results are provided for one typical two-group model problem to illustrate the accuracy of the numerical results and computer running time of the computer code implementing the proposed approach.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"221 ","pages":"Article 111540"},"PeriodicalIF":1.9,"publicationDate":"2025-05-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143936095","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":"Deployment of neural-network-based neutron microscopic cross sections in the Griffin reactor physics application","authors":"Olin W. Calvin ,&nbsp;Yifeng Che ,&nbsp;Yaqi Wang ,&nbsp;Paolo Balestra ,&nbsp;Javier Ortensi","doi":"10.1016/j.anucene.2025.111509","DOIUrl":"10.1016/j.anucene.2025.111509","url":null,"abstract":"<div><div>The use of reduced-order models for the efficient evaluation of microscopic cross sections has shown significant promise in Griffin reactor physics applications. Among various reduced-order techniques, neural network-based models stand out for their exceptional scalability, memory efficiency, prediction speed, and compatibility with the MOOSE (Multiphysics Object-Oriented Simulation Environment) framework. This work develops capabilities into the Griffin reactor physics application to utilize neural networks to predict microscopic cross-section parametric spaces for a variety of nuclides. The LibTorch interface enables Griffin’s MOOSE based materials to interact with LibTorch-trained models, allowing for the evaluation of complex microscopic cross-section spaces, which are then used to evaluate the neutronic properties of the Griffin finite-element model. This study benchmarks traditional ISOXML-formatted tabulation libraries against neural-network-based models for 279 nuclides on 20,160 grid points for zero- and two-dimensional reactor models. Benchmark metrics include the fundamental mode eigenvalue, fission and absorption rates, and various temperature coefficients of reactivity (isothermal, fuel, and moderator). From the perspective of storage space, the complete set of LibTorch models uses 11 MB, compared to the 10 GB for the ISOXML multigroup library that covers the same grid space. For the two-dimensional performance case considered in Griffin, the LibTorch model uses 97% less random access memory than the reference ISOXML dataset while runtime increases by a factor of 3 when using the LibTorch model compared to the ISOXML dataset with multilinear interpolation. The LibTorch model consistently yields errors within 0.01% for most analyzed quantities except for the temperature coefficients of reactivity where the maximum discrepancies are up to 0.3 <span><math><mfrac><mrow><mtext>pcm</mtext></mrow><mrow><mtext>K</mtext></mrow></mfrac></math></span>. Due to the neural network attempting to best predict quantities with no regard for a positive or negative bias for any given quantity, predictions may experience random fluctuations, resulting in both positive and negative errors. Future work will entail both a depletion and coupled transient analysis to determine the predictive capabilities of Griffin with neural-network-based cross sections.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111509"},"PeriodicalIF":1.9,"publicationDate":"2025-05-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143935914","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 novel viscoelastic creep model for zirconium alloy cladding under variable stress conditions","authors":"Chengjie Wang,&nbsp;Xiaoyan Wei,&nbsp;Yani Liu,&nbsp;Yayun Luo,&nbsp;Xin Jin","doi":"10.1016/j.anucene.2025.111551","DOIUrl":"10.1016/j.anucene.2025.111551","url":null,"abstract":"<div><div>Creep deformation in zirconium alloy cladding is a critical concern for reactor safety and reliability. However, the strain hardening rule is experimentally known to fail to describe the creep strain of zirconium alloy cladding in scenarios such as load drop or reversal. This study introduces a novel viscoelastic creep model based on the Boltzmann superposition principle, addressing the limitations of traditional time and strain hardening models under variable stress conditions. Validated against Halden Reactor Project experiments (IFA-585, IFA-663, IFA-699, IFA-741), the model demonstrates reduced complexity and slightly improved accuracy compared to the Tulkki model. Key parameters were derived using a least squares fitting algorithm, showing excellent agreement with experimental data. This model enhances predictive capabilities for cladding creep, suggesting potential improvements for fuel behavior codes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111551"},"PeriodicalIF":1.9,"publicationDate":"2025-05-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143935912","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":"Radionuclide calculation in the core components of Indonesia’s Triga 2000 reactor: Focus on tritium (H3) and its effect on decommissioning","authors":"Renaldy Bernardo Saragih ,&nbsp;Prasetyo Basuki ,&nbsp;Fajar Arianto ,&nbsp;Sherina Massayu Putri","doi":"10.1016/j.anucene.2025.111544","DOIUrl":"10.1016/j.anucene.2025.111544","url":null,"abstract":"<div><div>This paper provides estimations of the radionuclides created within the core components of the TRIGA 2000 Bandung reactor. This research is essential for the revision of decommissioning programs and forecasting the quantity and activity levels of radioactive wastes. This study predominantly focused on the tritium radionuclide within the reactor core. MCNP 6.1 was employed to determine the neutron flux during the reactor operation, while ORIGEN2.1 was utilized to identify the radionuclides produced as a result of neutron activation. The calculations extend across a nineteen-year, with 2020 as the base year. The modelling is focused on the calculation of the water components of the reactor pool. The findings suggest that Tritium is not the dominant isotope among all components from weight, with mass percentages almost zero but among the radionuclides, Tritium demonstrates a high level of activation and the highest level of radioactivity, ranging from 12% to 13% in the first year and declining over nineteen years as a result of radioactive decay. By the nineteenth year, a significant decrease in radioactivity was noted. Decommissioning nuclear facilities necessitates a comprehensive approach that considers factors such as cost, risk, potential environmental impacts, human safety, regulatory compliance, and the selected technical strategy.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111544"},"PeriodicalIF":1.9,"publicationDate":"2025-05-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143935915","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":"Main results of an in-vessel corium thermochemistry benchmark based on MASCA and CORDEB2 experimental data","authors":"R. Le Tellier ,&nbsp;L. Carénini ,&nbsp;F. Fichot ,&nbsp;R. Gencheva ,&nbsp;P. Groudev ,&nbsp;I. Melnikov ,&nbsp;M. Golubev ,&nbsp;A. Filippov","doi":"10.1016/j.anucene.2025.111514","DOIUrl":"10.1016/j.anucene.2025.111514","url":null,"abstract":"<div><div>This paper is dedicated to an international benchmark on in-vessel corium thermochemistry. Composed of three different steps, it was constructed in the frame of the IAEA Coordinated Research Project on In-Vessel Melt Retention and focuses on the stratification of corium in the lower plenum of a reactor vessel during a severe accident. The experimental data obtained from the MASCA and CORDEB2 programs is utilized to validate and calibrate the models of the various codes employed by the participants. The paper gives an overview of the different models, including their underlying assumptions. The results show that the models are generally in good agreement with the experimental data, but further experimental data is needed for model validation. Besides the need for further work, this paper highlights the fact that integral models (published in the open literature for most of them) are available to account for the first-order phenomena associated with in-vessel corium. Accordingly, this paper may be useful for code developers that would like to upgrade their models in order to take into account transient stratification kinetics, a prerequisite for a state-of-the-art analysis of in-vessel melt retention.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111514"},"PeriodicalIF":1.9,"publicationDate":"2025-05-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143935913","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 artificial neural network prediction of heat transfer in vertically upward internally rifled tubes with supercritical water","authors":"Wang Yibo ,&nbsp;Shao Huaishuang ,&nbsp;Wang Jingjie ,&nbsp;Shen Tao ,&nbsp;Liao Min ,&nbsp;Liang Zhiyuan ,&nbsp;Zhao Qinxin","doi":"10.1016/j.anucene.2025.111539","DOIUrl":"10.1016/j.anucene.2025.111539","url":null,"abstract":"<div><div>This paper presents an ANN model for heat transfer prediction of heat transfer in vertically upward internally rifled tubes with supercritical water. The model is trained using a dataset of 2071 experimental data points on supercritical heat transfer in internally rifled tubes. Dimensionless number of rib geometry associated with rib geometry are incorporated into the internally rifled tube structure, and optimization is performed on the input features and structure of the neural network. The obtained neural network model demonstrates accurate heat transfer predictions performance with overall test set <em>R²</em> = 0.9975 (coefficients of determination), <em>RMSE</em> = 0.2353 (root mean square error), and <em>MRE</em> = 1.65 % (mean relative error). The model has a prediction error range of 2 % for wall temperature, respectively. Furthermore, the neural network model exhibits superior accuracy in heat transfer predictions compared to several typical empirical correlations. This study provides a more extensive and accurate approach for predicting heat transfer in supercritical water.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111539"},"PeriodicalIF":1.9,"publicationDate":"2025-05-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143927887","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":"Verification of the reactor core sub-channel thermal hydraulic analysis code: CORTH under the flow transient conditions","authors":"Xiaoyu Wang ,&nbsp;Shinian Peng ,&nbsp;Jian Deng ,&nbsp;Yu Liu ,&nbsp;Sichao Tan ,&nbsp;Jiageng Wang ,&nbsp;Wei Zeng ,&nbsp;Xue Zhang","doi":"10.1016/j.anucene.2025.111538","DOIUrl":"10.1016/j.anucene.2025.111538","url":null,"abstract":"<div><div>The resistance characteristics of the reactor core will change under flow transient conditions. So, it is necessary to conduct verification and validation research on the core sub-channel code under flow transient conditions. Firstly, sub-channel analysis code CORTH is briefly introduced, and then CORTH is validated based on rod bundle pressure drop experiment. The verification and validation results indicate that the code is in good agreement with the experimental results. For the rod bundle channel bare rod area, the deviation between the calculated pressure drop and the experimental measured pressure drop is less than 6%. For the spacer grid area, the deviation between the calculated pressure drop and the measured pressure drop is mostly less than 8%. The validated CORTH can be used to analyze and calculate the resistance characteristics of the core under flow transient conditions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111538"},"PeriodicalIF":1.9,"publicationDate":"2025-05-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143928011","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":"Use of multinomial combinatorics for faster computations of neutron Thermal Scattering Law","authors":"R.S. Keshavamurthy ,&nbsp;Ananth S. Iyengar ,&nbsp;Mala N. Rao","doi":"10.1016/j.anucene.2025.111536","DOIUrl":"10.1016/j.anucene.2025.111536","url":null,"abstract":"<div><div>In this note, which builds upon our recently published work on an analytical series for the neutron Thermal Scattering Law (TSL), we employ combinatorial methods to reformulate the algebraic series into a more compact form with significantly fewer terms. We demonstrate that this reformulated series offers markedly improved computational efficiency, particularly when evaluating TSL for a large number of energy transfer (β) values. The approach shows good potential for efficient generation of large TSL data in reactor applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111536"},"PeriodicalIF":1.9,"publicationDate":"2025-05-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143922454","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 new network-level multi-objective optimization of the heat rejection system for megawatt space nuclear power systems","authors":"Yiheng Fei,&nbsp;Jianghan Fu,&nbsp;Chenglong Wang,&nbsp;Suizheng Qiu","doi":"10.1016/j.anucene.2025.111533","DOIUrl":"10.1016/j.anucene.2025.111533","url":null,"abstract":"<div><div>In Brayton, Stirling, or Rankine space nuclear systems, the Heat Rejection System (HJS) represents a critical component. The size and mass of the HJS significantly impact the overall performance of the space nuclear power system. This paper focuses on a three-wing HJS suitable for megawatt space nuclear power systems. The flow calculation model was coupled with the thermal radiation model to form the HJS calculation model, which was validated with an average error of 0.28 %. A novel network-level optimization method is proposed to optimize the mass density, pressure loss, and outlet temperature of the HJS by using flow connectivity and direction between different radiant units within the wings as independent variables. Optimal solutions for each loss function indicate that the outlet temperature has decreased from 390.70 K to 390.33 K, loop pressure loss has been reduced from 365 kPa to 361.38 kPa, and the mass density of the HJS has been reduced from 3.58 kg/m² to 3.33 kg/m². A simplified calculation method for the pre-design process of the HJS within the overall Space Nuclear Power System has been proposed and verified. This research provides valuable insights into the design of the HJS in megawatt-class space nuclear power systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111533"},"PeriodicalIF":1.9,"publicationDate":"2025-05-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143922372","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":"Wide wall liquid film thickness distribution in narrow rectangular geometry for annular flow","authors":"Peng Ju ,&nbsp;Xinhai Xu ,&nbsp;Zhengqian Qi ,&nbsp;Yi Qu ,&nbsp;Baixuan Jiang ,&nbsp;Peize Li ,&nbsp;Takashi Hibiki","doi":"10.1016/j.anucene.2025.111506","DOIUrl":"10.1016/j.anucene.2025.111506","url":null,"abstract":"<div><div>Annular flow in narrow rectangular geometries is critical for high-efficiency heat transfer systems, yet existing models—primarily developed for circular pipes—fail to capture the two-dimensional film thickness distribution inherent to asymmetric channels. This study addresses this gap by proposing a novel methodology that adapts conventional one-dimensional film thickness models to rectangular geometries through geometric and hydrodynamic considerations. Three characteristic lengths (hydraulic diameter, center-to-wall distance, and narrow gap) and two velocity profiles (1/7th power law turbulence and pressure-drop-derived phase separation) are systematically integrated into six methodologies to predict wide-wall film thickness. Validated against experimental data from a 200 mm × 10 mm rectangular channel, the results demonstrate that Method 6—combining hydraulic diameter and pressure-drop-derived superficial liquid velocity—achieves better accuracy by accounting for gas-core dominance at the channel center and confinement effects. While the framework reliably predicts wide-wall film thinning, challenges persist near the narrow wall due to flow regime transitions. This work advances the theoretical foundation for annular flow modeling in non-circular geometries, offering critical insights for enhancing thermal–hydraulic safety in nuclear reactors, compact heat exchangers, and high-power electronics cooling systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111506"},"PeriodicalIF":1.9,"publicationDate":"2025-05-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143918598","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}