Annals of Nuclear Energy最新文献

Transients analysis of the offshore fixed multi-purpose integrated all-natural-cycle small lead-cooled reactor with the FRTAC code

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-15 DOI: 10.1016/j.anucene.2025.111331

Zinan Huang , Yuxing Liu , Song Li , You-Peng Zhang , Lu Meng , Yunxiang Li , Wenjun Hu , Yang Yu , Xing-Kang Su , Lu Zhang

{"title":"Transients analysis of the offshore fixed multi-purpose integrated all-natural-cycle small lead-cooled reactor with the FRTAC code","authors":"Zinan Huang , Yuxing Liu , Song Li , You-Peng Zhang , Lu Meng , Yunxiang Li , Wenjun Hu , Yang Yu , Xing-Kang Su , Lu Zhang","doi":"10.1016/j.anucene.2025.111331","DOIUrl":"10.1016/j.anucene.2025.111331","url":null,"abstract":"<div><div>The offshore fixed multi-purpose integrated all-natural circulation small lead-cooled reactor is one of the safest and most economical offshore nuclear power plant solution. In order to verify the safety and rationality of the reactor design, this study used the lead-cooled reactor safety analysis code FRTAC to complete the construction of the offshore reactor loop model and simulated unprotected transient overpower accidents (UTOP) and unprotected Loss of Heat Sink accidents (ULOHS). The reactor physics parameters of the offshore reactor and the fission gas yield were calculated by the neutron transport calculation code Serpent. Based on the transient accident calculation results, we can preliminarily confirm that the offshore reactor can withstand the introduction of reactivity of up to 0.42$ under UTOP scenario; and under different degrees of ULOHS scenario, the maximum operating temperature of the offshore reactor core will not exceeds its failure limits.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111331"},"PeriodicalIF":1.9,"publicationDate":"2025-03-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143628490","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

MURR LEU structural and thermal hydraulics analyses: Part I – Preliminary irradiation thermo-mechanical behavior

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-15 DOI: 10.1016/j.anucene.2025.111350

F. Cetinbas , W. Mohamed , D. Yoon , J. Stillman , V. Mascolino , M. Pinilla , E. Wilson

{"title":"MURR LEU structural and thermal hydraulics analyses: Part I – Preliminary irradiation thermo-mechanical behavior","authors":"F. Cetinbas , W. Mohamed , D. Yoon , J. Stillman , V. Mascolino , M. Pinilla , E. Wilson","doi":"10.1016/j.anucene.2025.111350","DOIUrl":"10.1016/j.anucene.2025.111350","url":null,"abstract":"<div><div>The University of Missouri Research Reactor (MURR) is expected to be converted from highly enriched uranium (HEU, ≥ 20 wt% U-235) U-Al<sub>x</sub> dispersion fuel to low-enriched uranium (LEU, < 20 wt% U-235) with U-10Mo monolithic fuel. This work introduces high-fidelity irradiation thermo–mechanical (T-M) analysis of the MURR LEU focusing on changes in coolant channel gap thickness. Three-dimensional (3D) finite element (FE) models were developed to simulate the irradiation T-M behavior of the MURR LEU element with all 23 curved fuel plates, the two side plates, and the combs. It was shown that channel gap thickness changes were influenced not only by plate thickness variations due to fuel swelling and creep but also by the radial displacement of consecutive MURR LEU plates. Modeling the fuel element assembly captured side plate displacements, which were shown to reduce radial fuel plate displacements towards the convex side. The maximum local radial displacement in the element was predicted at the end of life (EOL) as 23.7 mil (602.0 µm) on the lateral centerline of plate 23 towards the convex side. The maximum stripe-averaged reduction in channel gap thickness, particularly relevant for thermal hydraulics (TH) safety analysis, was calculated as 15.9 mil (403.9 µm) in single-side heated channel 24 (the outermost channel). These results account for the thermal resistance from the oxide build-up on cladding surfaces which was shown to be up to 0.82 mil (20.8 µm) thick. It was demonstrated that accounting for oxide layer thermal resistance led to a 10 °C higher peak fuel temperature and a 4.4 mil (111.8 µm) greater maximum local radial displacement. The impact of the calculated channel gap thickness changes on the MURR LEU TH safety analysis is evaluated in Part II.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111350"},"PeriodicalIF":1.9,"publicationDate":"2025-03-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143629057","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

Study on the dynamic characteristics of hybrid ceramic ball bearings in a horizontal canned motor reactor coolant pump

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-14 DOI: 10.1016/j.anucene.2025.111314

Xiaohang Chen , Danrong Song , Jinqi Lu , Bo Wei , Rui Xu , Dezhong Wang

{"title":"Study on the dynamic characteristics of hybrid ceramic ball bearings in a horizontal canned motor reactor coolant pump","authors":"Xiaohang Chen , Danrong Song , Jinqi Lu , Bo Wei , Rui Xu , Dezhong Wang","doi":"10.1016/j.anucene.2025.111314","DOIUrl":"10.1016/j.anucene.2025.111314","url":null,"abstract":"<div><div>Horizontal Canned Motor Reactor Coolant Pump(HRCP) design attempts to use ceramic water-lubricated ball bearings. To demonstrate the suitability and safety of ceramic water-lubricated bearings, it is necessary to conduct an in-depth study of the dynamic characteristics under operating conditions. To improve the operational stability of the rotor and ensure that the bearings operate stably in water, we use hybrid ceramic ball bearings to support the rotor. However, the high-temperature environment in HRCP can affect the dynamic and support characteristics of the bearings, thereby impacting the safety of the rotor. Therefore, it is necessary to study the dynamic characteristics of hybrid ceramic bearings under high-temperature conditions. Based on the temperature field analysis of the HRCP, this paper establishes the load balance equations for hybrid ceramic deep groove ball bearings, incorporating the effects of thermal expansion. This analysis model was used to systematically study the contact angles and dynamic characteristics of steel ball bearings and ceramic ball bearings under different temperatures, speeds, and loads. The results indicate that the stiffness of ceramic ball bearings is higher than that of steel ball bearings. Additionally, as the temperature and speed increase, the stiffness of the bearings decreases.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111314"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621257","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

Potential of deep learning methods to enhance satellite-based monitoring of nuclear power plants focusing on remote operation evaluations

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-14 DOI: 10.1016/j.anucene.2025.111337

Hui-Yu Hsieh , Thabit Abuqudaira , Pavel Tsvetkov , Piyush Sabharwall

{"title":"Potential of deep learning methods to enhance satellite-based monitoring of nuclear power plants focusing on remote operation evaluations","authors":"Hui-Yu Hsieh , Thabit Abuqudaira , Pavel Tsvetkov , Piyush Sabharwall","doi":"10.1016/j.anucene.2025.111337","DOIUrl":"10.1016/j.anucene.2025.111337","url":null,"abstract":"<div><div>The anticipated expansion of the nuclear industry and the deployment of new nuclear reactors (200 + GW of new nuclear capacity by 2050) require the development of monitoring systems that align with safety and security concerns, providing enhanced evaluation capabilities. A remote monitoring system using satellites and deep learning techniques was evaluated for its ability to detect anomalies and capture various features of nuclear reactors independently of the conditions on the ground. Satellite images of current operational and under-construction nuclear power plants were collected from Google Earth Pro as a surrogate database. Subsequently, five datasets were created from the collected images. Transfer learning technique was used for several classification tasks utilizing VGG16, ResNet50V2, Xception, DenseNet121, and MobileNetV2 pre-trained models. In the first task, the capability of the monitoring system to detect abnormal conditions or processes in a nuclear power plant was investigated. In the second task, the ability to capture operational features remotely was examined. As an example, for the purposes of this study, these features included classifying reactors based on type, power range, or onsite condition. Several evaluation metrics were used to compare the performance of the pre-trained models and the overall monitoring system. The evaluation results demonstrated that deep learning techniques and pre-trained models applied to satellite images have the potential to facilitate further and expand capabilities in monitoring systems to assess plant operation details.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111337"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621258","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

Numerical study on gas–liquid separation of two-phase swirling flow based on the Eulerian-Eulerian approach and RSM turbulence model

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-14 DOI: 10.1016/j.anucene.2025.111334

Qian Zhang , Wenzhen Chen , Hu Liu

{"title":"Numerical study on gas–liquid separation of two-phase swirling flow based on the Eulerian-Eulerian approach and RSM turbulence model","authors":"Qian Zhang , Wenzhen Chen , Hu Liu","doi":"10.1016/j.anucene.2025.111334","DOIUrl":"10.1016/j.anucene.2025.111334","url":null,"abstract":"<div><div>The steam-water separation package in U-tube steam generator is designed to minimize water content to ensure turbine safety. However, the design of this key component still relies on experiments and trial-and-error methods due to the lack of comprehensive analytical success. The two-phase flow inside an axial swirler is modeled by the Eulerian-Eulerian multiphase flow approach, in which both gas and liquid phases are considered as continuous and coupled to each other. The choice of an appropriate turbulence model significantly impacts on the accuracy of simulation results. The simulation performance of three advanced swirling turbulence models on the gas–liquid interface downstream of the swirler is analyzed. The void fractions along the axial cylinder sections above the swirl vanes are compared with the corresponding experimental air core diameter data from the literature. It is found that the void fraction calculated with RNG <em>k-ε</em> and Realizable <em>k-ε</em> turbulence models decreases monotonously with the axial height above the swirl vanes, whereas the RSM model shows an initial decrease followed by an increase. The RSM model’s results align well with experimental data, particularly in predicting the turning point of void fraction. Adequate grid coverage in the boundary layer is essential to capture the gas–liquid interface due to the very thin liquid film. To obtain the better prediction of gas–liquid swirl flow, the RSM turbulence model and appropriate boundary grid coverage are recommended.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111334"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621256","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

A linearization method for the transverse-leakage terms in hexagonal nodal method based on the conformal mapping technique

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-13 DOI: 10.1016/j.anucene.2025.111325

Chenghui Wan , Haozhe Yang , Jiahe Bai , Jianfu Zhang , Songzhe Wang , Wei Shen

{"title":"A linearization method for the transverse-leakage terms in hexagonal nodal method based on the conformal mapping technique","authors":"Chenghui Wan , Haozhe Yang , Jiahe Bai , Jianfu Zhang , Songzhe Wang , 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}

引用次数: 0

A nodal analytical discrete ordinates solver for pin-homogenized core calculation

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-13 DOI: 10.1016/j.anucene.2025.111311

Dean Wang

{"title":"A nodal analytical discrete ordinates solver for pin-homogenized core calculation","authors":"Dean Wang","doi":"10.1016/j.anucene.2025.111311","DOIUrl":"10.1016/j.anucene.2025.111311","url":null,"abstract":"<div><div>Advanced reactor designs require highly accurate and efficient neutronics modeling capabilities for practical design needs. Many transport codes are designed for high-fidelity pin-resolved calculations, which are often computationally expensive, while low order solvers such as those based on diffusion theory or SP<sub>N</sub> methods cannot accurately model transport effects and their solution is often prone to oscillations in regions with significant local heterogeneities. In this paper, we try to address such modeling challenges by developing a viable pin-wise whole-core transport solver based on a newly developed nodal analytical discrete ordinates (ANDO) method (Rocheleau and Wang, 2020, 2022). Here we extend the ANDO solution for fixed-source problems to k-eigenvalue problems, which are solved with the power iteration algorithm. In our implementation of the ANDO method, all the fission or part of it is combined with the scattering source and integrated analytically during each power iteration. Such a novel treatment can greatly improve computational accuracy and efficiency of the ANDO method for k-eigenvalue neutron transport calculations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111311"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621255","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

Sensitivity analysis of internal flow distribution in the Tsinghua high flux reactor

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-13 DOI: 10.1016/j.anucene.2025.111352

Yuan Huang, Meng Lv, Heng Xie, Lei Shi

{"title":"Sensitivity analysis of internal flow distribution in the Tsinghua high flux reactor","authors":"Yuan Huang, Meng Lv, Heng Xie, 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}

引用次数: 0

Experimental investigation of nonisothermal interaction between Fe-Zr melt and stainless steel forming “metallic debris” in Fukushima Daiichi Nuclear Power Station

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-12 DOI: 10.1016/j.anucene.2025.111333

Ayumi Itoh , Tatsuya Kanno , Takayuki Iwama , Shigeru Ueda , Takumi Sato , Yuji Nagae

{"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 , Tatsuya Kanno , Takayuki Iwama , Shigeru Ueda , Takumi Sato , 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 M<sub>23</sub>Zr<sub>6</sub>, M<sub>2</sub>Zr, and MZr<sub>2</sub> (M = Fe, Cr, Ni) was confirmed, with varying Ni concentrations in M<sub>23</sub>Zr<sub>6</sub> 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 FeCr<sub>2</sub>O<sub>4</sub> 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}

引用次数: 0

Development of lumped-parameter models for debris bed remelting analysis

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-03-11 DOI: 10.1016/j.anucene.2025.111348

Lijun Jian , Peng Yu , Xiao Zeng , Liangxing Li , Rubing Ma , Yidan Yuan , Weimin Ma

{"title":"Development of lumped-parameter models for debris bed remelting analysis","authors":"Lijun Jian , Peng Yu , Xiao Zeng , Liangxing Li , Rubing Ma , Yidan Yuan , 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}

引用次数: 0