Annals of Nuclear Energy最新文献

{"title":"A linearization method for the transverse-leakage terms in hexagonal nodal method based on the conformal mapping technique","authors":"Chenghui Wan ,&nbsp;Haozhe Yang ,&nbsp;Jiahe Bai ,&nbsp;Jianfu Zhang ,&nbsp;Songzhe Wang ,&nbsp;Wei Shen","doi":"10.1016/j.anucene.2025.111325","DOIUrl":"10.1016/j.anucene.2025.111325","url":null,"abstract":"<div><div>Widely used in the hexagonal-assembly core-analysis code, the conformal mapping technique has proved to be suitable, accurate, and efficient. Throughout the years of its fledging development, there was hardly any treatment generally applicable for the conformally mapped transverse-leakage terms. This issue notably affected the calculation accuracy of the hexagonal nodal calculation. To address this issue, in the present study, a linearization method for the transverse-leakage terms has been proposed, which estimates the current distribution of nodal surfaces with corresponding flux distribution on surfaces adjacent to neighboring nodes. This method provides an accurate distribution of the transverse-leakage terms, leading to calculation results with high accuracy.</div><div>The proposed method has been implemented in our in-house core-analysis code, SPARK, enabling the solution of the three-dimensional multi-group neutron-diffusion equation using hexagonal nodes.</div><div>To verify the method, the two-dimensional VVER-1000 benchmark problem was calculated in the first place. Compared with the conventional flat-current assumption, the proposed linearization method decreased the error of eigenvalue and the maximum error of the nodal normalized power from 62.9 pcm to 8.6 pcm and from 5.60% to −0.65%, respectively. Subsequently, numerous 2D/3D benchmarks were modeled and verified, comparing the eigenvalues and assembly-averaged power distributions with their corresponding reference values. The numerical results indicate that the proposed linearization method performs satisfactorily, reducing the maximum error in eigenvalue to about 20.0 pcm and keeping the errors in power distribution below 0.9%. As a result, the proposed linearization method significantly improves computation accuracy and offers an effective solution for handling the transverse-leakage terms using the conformal mapping technique.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111325"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610880","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":"Sensitivity analysis of internal flow distribution in the Tsinghua high flux reactor","authors":"Yuan Huang,&nbsp;Meng Lv,&nbsp;Heng Xie,&nbsp;Lei Shi","doi":"10.1016/j.anucene.2025.111352","DOIUrl":"10.1016/j.anucene.2025.111352","url":null,"abstract":"<div><div>This paper presents a simple approach for rapidly analyzing the flow distribution characteristics and sensitivity within the THFR (<strong>T</strong>singhua <strong>H</strong>igh <strong>F</strong>lux <strong>R</strong>eactor). Given the high flow rates and large mixing space provided by the upper and lower chambers in the high flux reactor, the flow network theory aligns well with these assumptions, resulting in good agreement between the theoretical calculations and CFD simulations. Based on the flow network theory, the sensitivity analysis of the internal structural flow areas within the reactor indicates that changes in the flow channel area typically dominate the impact; variations in the flow area of the larger control drums and external irradiation boxes have a more significant overall effect. In subsequent design iterations, adjusting the flow resistance distribution ratio among the branches can mitigate the impact of potential dimensional changes on the overall flow distribution. Meanwhile, variations in channel dimensions significantly alter the pressure drop, imposing higher demands on the pumps.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111352"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610881","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":"Experimental investigation of nonisothermal interaction between Fe-Zr melt and stainless steel forming “metallic debris” in Fukushima Daiichi Nuclear Power Station","authors":"Ayumi Itoh ,&nbsp;Tatsuya Kanno ,&nbsp;Takayuki Iwama ,&nbsp;Shigeru Ueda ,&nbsp;Takumi Sato ,&nbsp;Yuji Nagae","doi":"10.1016/j.anucene.2025.111333","DOIUrl":"10.1016/j.anucene.2025.111333","url":null,"abstract":"<div><div>In the Fukushima Daiichi Nuclear Power Station Unit 2, the formation of a metallic pool, mainly comprising Fe and Zr, has been proposed as a mechanism contributing to the failure of the reactor pressure vessel. This study focuses on material interactions during the early core degradation that led to metallic pool formation in the late phase of the in-vessel degradation process. It investigates the nonisothermal reaction between the Fe-Zr melt and stainless steel (SS), hypothesizing that metallic debris could have formed during the relocation of the melt along the SS structure to the lower region. Initially, two compositions, Fe-87Zr and Fe-15Zr (at%), were heated to the liquidus temperature of 1723 K, dropped onto SS at lower temperatures, and the metallographic structure of the reaction products was examined. The formation of intermetallic compounds such as M₂₃Zr₆, M₂Zr, and MZr₂ (M = Fe, Cr, Ni) was confirmed, with varying Ni concentrations in M₂₃Zr₆ depending on the Zr concentration of the melt. Subsequently, the Fe-87Zr melt at temperatures ranging from 1723 to 1873 K was dropped onto oxidized SS to evaluate the influence of the oxide layer on degradation. The oxide layer provided some protection to the degradation of SS; however, the Zr-rich melt corroded the FeCr₂O₄ oxide layer, 20 µm thick, above 1723 K, and severe degradation of SS was observed at 1873 K. In contrast, the Fe-rich melt did not react with the oxide layer due to poor wettability. This study confirmed that the liquidus temperatures of all intermetallic compounds were below 2000 K, and the metallic debris could be a source of the “metallic pool formation” predicted by recent severe accident analysis.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111333"},"PeriodicalIF":1.9,"publicationDate":"2025-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143611330","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":"Development of lumped-parameter models for debris bed remelting analysis","authors":"Lijun Jian ,&nbsp;Peng Yu ,&nbsp;Xiao Zeng ,&nbsp;Liangxing Li ,&nbsp;Rubing Ma ,&nbsp;Yidan Yuan ,&nbsp;Weimin Ma","doi":"10.1016/j.anucene.2025.111348","DOIUrl":"10.1016/j.anucene.2025.111348","url":null,"abstract":"<div><div>During postulated severe accidents of a light water reactor, a debris bed may form in the lower head of the reactor pressure vessel due to Fuel-Coolant Interaction (FCI), and re-melt into a molten pool if the debris bed is uncoolable. The debris bed remelting is therefore an important process in a severe accident scenario. To predict the dynamic process of debris bed remelting, a computer program is developed in the present study using lumped-parameter models. The melt in the lower head is split into different zones of molten metal, molten oxide and solid debris particles submerged in molten pools. Correlations are employed to calculate the heat transfer within each zone and between zones. The developed lumped-parameter code is employed to calculate the COREM experiments. The comparison of the simulation results with the experimental shows a reasonable agreement for melting processes of single-material and two-material debris beds. The code is also used to investigate some factors which may affect debris bed remelting, such as internal heating power, volume ratio of components, and thermophysical properties.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111348"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592916","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 theoretical design procedure for cladding protecting lower head of central measuring shroud from thermal shock damage","authors":"Shu Zheng ,&nbsp;Daogang Lu ,&nbsp;Qiong Cao ,&nbsp;Yuxiong Xue","doi":"10.1016/j.anucene.2025.111338","DOIUrl":"10.1016/j.anucene.2025.111338","url":null,"abstract":"<div><div>The cladding serves as a protective barrier for the central measuring shroud, safeguarding it from thermal shock damage caused by the SCRAM events. To facilitate rapid preliminary design, a theoretical design procedure of the cladding was developed based on thermal, mechanical and creep-fatigue damage theories. Then, the design was performed according to actual operating conditions. It was found that the procedure can reduce design time and computational costs of the design, but needs to be adjusted because of stress concentration, with an adjustment factor of 7.32 for the total thickness design and 22.677 for the layer thickness design. The final design features a total cladding thickness of 6 mm, comprising two layers of 3 mm each. Analysis showed that cladding can mitigate heat conduction from the coolant. Specifically, increasing the cladding thickness from 0 to 6 mm reduced the maximum temperature difference by 49 °C and decreased the maximum stress amplitude by 2.35 × 10⁸ Pa.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111338"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592915","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 coordinate control strategy for load following operation of sodium-cooled fast reactor system","authors":"Qingfeng Jiang ,&nbsp;Jinrong Jin ,&nbsp;Areai Nuerlan ,&nbsp;Yizhe Liu ,&nbsp;Jiashuang Wan ,&nbsp;Pengfei Wang","doi":"10.1016/j.anucene.2025.111318","DOIUrl":"10.1016/j.anucene.2025.111318","url":null,"abstract":"<div><div>The sodium-cooled fast reactor (SFR), as one of the most mature and promising Generation IV reactors, is a potential energy source. Load following operation has flexible regulation performance and is suitable for future development trend of grid peaking operation. The paper proposes a coordinate control strategy for the SFR under load following operation mode. Firstly, a simulation platform for the SFR was developed by adopting MATLAB/Simulink library technology. Secondly, a coordinate control strategy under load following operation mode was developed to coordinate primary sodium loop, secondary sodium loop, and feedwater-steam loop. Finally, typical operational transients such as step and ramp load change transients were simulated to validate the developed coordinate control strategy. The simulation results demonstrate that the reactor power has satisfactory flexibility and the system key parameters have good dynamic characteristics, which ensures expected performances of the SFR.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111318"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143593025","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":"SFR core-catcher safety issue: Experimentation and simulation of its ablation by an impinging jet","authors":"A. Avrit ,&nbsp;N. Seiler ,&nbsp;A. Lecoanet ,&nbsp;A. Djermoune ,&nbsp;N. Rimbert ,&nbsp;M. Gradeck","doi":"10.1016/j.anucene.2025.111329","DOIUrl":"10.1016/j.anucene.2025.111329","url":null,"abstract":"<div><div>These coupled experimental and numerical studies support the safety demonstration of dedicated mitigation features in Sodium Fast Reactors (SFR). In order to limit potential power excursion in SFR core in the case of hypothetical severe accident, the implementation of the mitigation devices, called transfer tubes, was considered into actual French SFR core design. More specifically, they connect the core to the lower plenum and enable the early discharge of the corium towards the core-catcher. However, these devices, which effectively reduce the probability of reactivity excursions in the core, emphasize the issue of core-catcher thermomechanical resistance. Contrary to previous reactor core designs, the presence of these tubes could lead to coherent corium jets impinging the core-catcher surface. In the past, numerous studies dedicated to thermal ablation of a solid by a jet have been carried out to characterize the maximum ablated depth. Although these experiments are very valuable at a macroscopic scale, they do not give information about the local physical phenomena governing heat transfer and melting. Keeping this final goal in mind, this study followed a R&amp;D methodology based on an experimental situation using simulant materials enabling direct visualization. In this paper, the ablation of an ice block by an immersed water jet and a free surface jet are investigated in conditions shown to be close to the reactor case. The physical analysis of the heat transfer at jet impingement and in its vicinity is characterized for the immersed jet conditions. Associated to the experimental part, CFD simulations of these tests are performed to validate the calculation methodology (mesh, turbulence models and interface melting…). In the future, simulations will be used in particular to characterize steel/ice transposition biases.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111329"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592917","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":"A study of the application of HFETR and its safety review management principles","authors":"Liang Zhang,&nbsp;Qibing Chen,&nbsp;Mingyan Tong,&nbsp;Guoqiang Liu,&nbsp;Peng Liu,&nbsp;Panqing He","doi":"10.1016/j.anucene.2025.111351","DOIUrl":"10.1016/j.anucene.2025.111351","url":null,"abstract":"<div><div>With the rapid development of national economy and scientific and technological capabilities over the past decades, irradiation test and isotope production of research reactors have grown considerably, and the associated safety risks and controls accompanying the research reactor operations and applications have become increasing critical. With the gradual diversification of irradiation test methods and test materials in research reactors, it is difficult for the relevant regulations and guidance frameworks based on HAF/HAD (nuclear safety laws and regulations of China / nuclear safety guidance of China) to cover all types of applications, creating challenges in safety review and management for research reactor applications in practice. In this study, we have sorted out the relevant laws, regulations and bases of research reactor application safety review, categorized the research reactor applications and historical data of HFETR (High Flux Engineering Test Reactor), analyzed the specific application safety review practices of HFETR, and proposed the management principles for its application safety review. The research findings presented in this paper can provide technical references for safety review processes and management practices pertaining to applications in similar research reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111351"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579679","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":"Mapping radionuclide concentrations in the UAE using a Gaussian process Machine learning approach","authors":"Bassam A. Khuwaileh ,&nbsp;Belal Almomani ,&nbsp;Samar El-Sayed ,&nbsp;Rahaf Ajaj ,&nbsp;Yumna Akram","doi":"10.1016/j.anucene.2025.111335","DOIUrl":"10.1016/j.anucene.2025.111335","url":null,"abstract":"<div><div>This study introduces a machine learning approach for mapping soil radionuclide concentrations in the UAE using Gaussian Process (GP) regression. Aimed at enhancing environmental monitoring and public health, the approach utilizes soil samples from across the UAE to create a radiation map (covering Ra-226, Th-232, and K-40 isotopes). GP regression, known for its proficiency in spatial data interpolation, predicts radionuclide levels in areas without direct testing. This method is adept at managing the non-linear spatial complexities inherent in geographic data, offering both a qualitative and quantitative understanding beneficial for decision-making and further sampling strategies. The results reveal the GP model’s capacity to accurately reflect geographic variances in isotopic concentrations, with RMSE values of 13%, 14%, and 22% for Ra-226, Th-232, and K-40, respectively. The model’s success in learning the geographical variations of the concentrations showcases its potential to guide future research by identifying areas of increased radioactivity risk.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111335"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579677","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":"Neutronic analysis for withdrawal of TRIGA MARK II reactor control rods using OpenMC program","authors":"Fajri Prasetya ,&nbsp;Zaki Su’ud ,&nbsp;Nuri Trianti ,&nbsp;Efrizon Umar ,&nbsp;Ratna Dewi Syarifah","doi":"10.1016/j.anucene.2025.111330","DOIUrl":"10.1016/j.anucene.2025.111330","url":null,"abstract":"<div><div>Currently, the TRIGA 2000 Bandung reactor which is a MARK II type has reshuffling some of the fuel elements to maintain its criticality during operation. The reshuffling includes filling and moving fuel which has a different uranium composition for each element. Heterogeneity in the composition of fuel elements has the potential to produce fairly high neutron flux and PPF (Power Peaking Factor) values. This is the main focus that needs to be discussed in this study. The purpose of this study is to conduct a neutronic analysis in the form of calculating the values of k-eff, PPF, neutron flux spectrum, reactions rate, neutron flux distribution, and power density distribution with several control rod withdrawal conditions. Analysis of two types of PPF is also discussed, including APF (Axial Power Peaking Factor) and RPF (Radial Power Peaking Factor). Validation in the form of k-eff and APF parameters of the F7 fuel element has been carried out to see the accuracy of the OpenMC code. The results of the validation of the OpenMC and MCNP neutronic parameters for all control rod withdrawal variations show accurate value agreement with %Δk &lt; 1 %. The results of the k-eff value and excess reactivity indicate that the reactor can operate critically if the control rod withdrawal percentage is &gt;50 %. Increasing the control rod percentage significantly affects the APF and RPF value, which shifts and decreases. At 60 % and 100 % withdrawal, the maximum APF values are 1.332 and 1.245. Meanwhile, the RPF values are 2.042 and 1.898. The withdrawal of control rods has a slight impact on the increase in neutron flux spectrum and the fission-absorption reaction rate in the fuel, particularly at thermal energies. A different condition occurs for the maximum neutron flux in the inner radial ring, which decreases as the control rods are withdrawn. The power density distribution results of the 111 fuel elements show that the reactor core still has heterogeneity. This presents a prospect for future research in designing the TRIGA 2000 Bandung reactor to achieve a more uniform power distribution.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111330"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143593024","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}