Annals of Nuclear Energy最新文献

{"title":"Parallel rCMFD acceleration method for the Triangular-Z nodal transport calculation of fast reactors","authors":"Zhi-Tao Xu,&nbsp;Wei-Guang Li,&nbsp;Ai-Jun Niu","doi":"10.1016/j.anucene.2024.110994","DOIUrl":"10.1016/j.anucene.2024.110994","url":null,"abstract":"<div><div>The discrete nodal transport method (DNTM) in triangular-Z meshes is favored for fast reactor calculations due to its precision, efficiency, and geometric flexibility. However, the computational demands of frequent core calculations necessitate efficient methods. Based on the framework of the triangular-Z DNTM, this paper adapts the effective revised coarse-mesh finite difference (rCMFD) acceleration method to triangular prisms and explores MPI parallelism for transport and diffusion with spatial domain decomposition. We verify and evaluate the proposed parallel rCMFD method based on three fast reactor problems: Takeda-3, Takeda-4, and MOX-3600. Results show negligible impact on key parameters like <em>k</em>_eff and flux distribution, but significant speedups. Under single-core, rCMFD achieved speedups of 2.5, 3.6, and 6.2 for the three problems. For100-core scenario, without rCMFD, parallel speedups peaked at 142.9–210.7, while with rCMFD, total speedups reached 262.5–669.8, resulting in minimal computing times of 9.0 s, 1.1 s, and 61.8 s, respectively.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142447005","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 a conservative safety demonstration for loss of forced flow events in MYRRHA","authors":"Tewfik Hamidouche,&nbsp;Guy Scheveneels,&nbsp;Rafaël Fernandez","doi":"10.1016/j.anucene.2024.110981","DOIUrl":"10.1016/j.anucene.2024.110981","url":null,"abstract":"<div><div>The MYRRHA facility is a fast research reactor cooled by lead–bismuth eutectic (LBE). It is designed to operate in both subcritical and critical modes. In subcritical mode, coupled with an accelerator, it will demonstrate the full concept of Accelerator Driven Systems at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allows for the study of efficient transmutation of high-level nuclear waste.</div><div>In this paper, an example of a robust safety demonstration covering all initiating events related to primary pump failures is given. The different steps to determine an enveloping event and to construct a conservative evaluation model are depicted. The analysis shows that even for a very conservative postulated event there is considerable safety margin, and no pump related failures can jeopardize the integrity of the cladding barrier.</div><div>The analysis is performed using a model set up in the RELAP5/3D v4.3.4 system thermal–hydraulic (STH) code. The analysis shows that for the MYRRHA reactor there is ample of time available to scram the reactor and prevent the cladding failure. Since there is no cladding failure the radiological safety objective 1 for category C-2 (anticipated operational occurrences) is met for all primary pump related failure initiating events. Sensitivity analyses were performed to identify parameters that have an important impact on the loss of forced flow transients. The result is that all important parameters have sufficient margin compared to their uncertainty.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142447006","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":"Data assimilation of turbulent flow in a large-scale steam generator: Part I- Iterative ensemble-Kalman filter-based reconstruction","authors":"Sen Li ,&nbsp;Yuheng Lu ,&nbsp;Chuangxin He ,&nbsp;Chunjing Song ,&nbsp;Yingzheng Liu ,&nbsp;Yun Zhong","doi":"10.1016/j.anucene.2024.110982","DOIUrl":"10.1016/j.anucene.2024.110982","url":null,"abstract":"<div><div>This research focuses on reproducing the global turbulent mean flow within a large-scale steam generator (SG) system using an iterative Ensemble Kalman Filter (EnKF)-based data assimilation (DA). A compressed directional loss model is introduced to reduce computational costs while considering volume flow rate redistribution across the U-shaped arrays. Results demonstrate that the DA approach improves predictions, showing better agreement with experimental data by widening the jet core, enhancing jet array penetration, and reducing turbulent separation bubble size. The inlet velocity profile at the reactor coolant pump (RCP) entrance is also accurately represented. The extensibility of the optimized model is validated at the RCP outlet. The DA model more accurately captures fluid dynamics, including acceleration, deceleration, and vertical movement in the sudden expansion region, leading to better estimations of total pressure loss. These improvements open up possibilities of DA approach for real engineering applications in both design and operation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142442455","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":"Broomstick experiment with copper in VR-1 reactor","authors":"Michal Košťál ,&nbsp;Evžen Losa ,&nbsp;Tomáš Czakoj ,&nbsp;Stanislav Simakov ,&nbsp;Marek Zmeškal ,&nbsp;Martin Schulc ,&nbsp;Jan Šimon ,&nbsp;Vojtěch Rypar ,&nbsp;Evžen Novák ,&nbsp;František Cvachovec ,&nbsp;Filip Mravec ,&nbsp;Václav Přenosil ,&nbsp;Peter Krásný ,&nbsp;Roberto Capote ,&nbsp;Zdeněk Matěj","doi":"10.1016/j.anucene.2024.110993","DOIUrl":"10.1016/j.anucene.2024.110993","url":null,"abstract":"<div><div>Copper is an important structural material used in nuclear technology, often used as a<!--> <!-->cover for spent fuel canisters or planned to be used in fusion devices. Despite its significance, there is a lack of integral experiments useful for validating and improving the evaluations of copper nuclear data. To address this gap, a neutron leakage experiment was conducted a few years ago using a point ²⁵²Cf(s.f.) neutron source placed inside a large block of copper. In this work a pencil beam transmission-attenuation experiment (a broomstick) employing various thicknesses (5–20 cm) of copper blocks (cylinders of 6 cm in diameter) was undertaken to expand the dataset of available experiments for copper in the fast neutron energy range (1–10 MeV). This type of experiment has the highest sensitivity to the total cross sections, and sensitivities are different from other integral experiments, making it a complementary measurement to already existing integral data. The measurement was performed using stilbene scintillation spectrometry. Measured transmission shows that the current INDEN evaluation, proposed to be adopted for ENDF/B-VIII.1 and JEFF-4 libraries, exhibits excellent agreement with experimental data. The JEFF-3.3 evaluation displays significant discrepancies, consistent with previous results from integral experiments involving copper. In the case of JENDL-5, discrepancies were found in the energy region 1.7–4.9 MeV.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142442456","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":"Preliminary exploration of liquid metals turbulent heat flux model based on OpenFOAM solver: Second-order differential heat flux model","authors":"Yuefeng Guo ,&nbsp;Xingkang Su ,&nbsp;Guan Wang ,&nbsp;Long Gu ,&nbsp;Xianwen Li","doi":"10.1016/j.anucene.2024.110969","DOIUrl":"10.1016/j.anucene.2024.110969","url":null,"abstract":"<div><div>Liquid metal-cooled fast reactors use liquid metals such as lead–bismuth eutectic (LBE) and sodium as the coolant, and the thermo-hydraulic characteristics of liquid metals have a large effect on the thermodynamic parameters of the reactor core. The problem is that using the traditional constant turbulent Prandtl number <em>Pr_t</em> of 0.85 ∼ 0.9, which is generally derived from the Reynolds analogy hypothesis suitable for conventional fluids, will greatly affect the accuracy of the numerical prediction of the thermal–hydraulic properties of liquid metals. In order to obtain more accurate liquid LBE and sodium turbulent heat transfer data, this research introduces a second-order differential heat flux model (DHFM) based on the open-source computational fluid dynamics platform OpenFOAM. The turbulent heat flux model is also tested with geometrical model of a pipe flow, the square and triangular bundle flow. It is shown that the calculation of Nusselt number <em>Nu</em> for liquid LBE by the second-order differential heat flux model (DHFM) is larger than the correlations in the square and triangular bundle flow. While the calculated <em>Nu</em> for liquid sodium is smaller than the correlations in the square and triangular bundle flow. Among them, the calculated results of liquid LBE and sodium in the pipe flow are in good agreement with the experimental data and the correlations. The DHFM model has a larger error in the square and triangular bundle channels for liquid LBE and sodium. In addition, the non-dimensional temperature, temperature fluctuation and dissipation characteristics of different geometrical models are also investigated in this research. This research can serve as a reference for the numerical calculation of turbulent heat transfer in liquid metals. It also enriches the study of thermal-hydraulics in liquid metal reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142437737","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 multi-scale coupling algorithm for burnup calculation of dispersed particulate poison","authors":"Junxian Li ,&nbsp;Xuezhong Li ,&nbsp;Jiejin Cai","doi":"10.1016/j.anucene.2024.110980","DOIUrl":"10.1016/j.anucene.2024.110980","url":null,"abstract":"<div><div>The spatial effects of dispersed particulate medium have a significant impact on the neutron characteristics of poison, leading to pronounced microscopic stratification phenomenon during burnup. A new multi-scale coupling burnup calculation algorithm for the dispersed particulate poison has been proposed. The microscopic exact particle model is coupled with the macroscopic homogeneous lattice element model. This coupling simplifies the direct refined solution process of dispersed particulate medium into a fast-solving problem for a simple and conventional medium. The computation results show that the multi-scale coupling algorithm has an overall error of effective multiplication factor around 200 pcm. The error level of thermal neutron flux and poison number density by using new algorithm are 1.2% and 1.41%, respectively. The overall time for calculation of each burnup step is twice as efficient as the direct-refined method.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142432199","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":"Uncertainty quantification for severe-accident reactor modelling: Results and conclusions of the MUSA reactor applications work package","authors":"S. Brumm ,&nbsp;F. Gabrielli ,&nbsp;V. Sanchez Espinoza ,&nbsp;A. Stakhanova ,&nbsp;P. Groudev ,&nbsp;P. Petrova ,&nbsp;P. Vryashkova ,&nbsp;P. Ou ,&nbsp;W. Zhang ,&nbsp;A. Malkhasyan ,&nbsp;L.E. Herranz ,&nbsp;R. Iglesias Ferrer ,&nbsp;M. Angelucci ,&nbsp;M. Berdaï ,&nbsp;F. Mascari ,&nbsp;G. Agnello ,&nbsp;O. Sevbo ,&nbsp;A. Iskra ,&nbsp;V. Martinez Quiroga ,&nbsp;M. Nudi ,&nbsp;T. Sevon","doi":"10.1016/j.anucene.2024.110962","DOIUrl":"10.1016/j.anucene.2024.110962","url":null,"abstract":"<div><div>The recently completed Horizon-2020 project “Management and Uncertainties of Severe Accidents (MUSA)” has reviewed uncertainty sources and Uncertainty Quantification methodology for assessing Severe Accidents (SA), and has made a substantial effort at stimulating uncertainty applications in predicting the radiological Source Term of reactor and Spent Fuel Pool accident scenarios.</div><div>The key motivation of the project has been to bring the advantages of the Best Estimate Plus Uncertainty approach to the field of Severe Accident modelling. With respect to deterministic analyses, expected gains are avoiding adopting conservative assumptions, identifying uncertainty bands of estimates, and gaining insights into dominating uncertain parameters. Also, the benefits for understanding and improving Accident Management were to be explored.</div><div>The reactor applications brought together a large group of participants that set out to apply uncertainty analysis (UA) within their field of SA modelling expertise – in particular reactor types, but also SA code used (ASTEC, MELCOR, MAAP, RELAP/SCDAPSIM), uncertainty quantification tools used (DAKOTA, SUSA, URANIE, self-developed tools based on Python code), detailed accident scenarios, and in some cases SAM actions. The setting up of the analyses, challenges faced during that phase, and solutions explored, are described in Brumm et al. ANE 191 (2023).</div><div>This paper synthesizes the reactor-application work at the end of the project. Analyses of 23 partners are presented in different categories, depending on whether their main goal is/are (i) uncertainty bands of simulation results; (ii) the understanding of dominating uncertainties in specific sub-models of the SA code; (iii) improving the understanding of specific accident scenarios, with or without the application of SAM actions; or, (iv) a demonstration of the tools used and developed, and of the capability to carry out an uncertainty analysis in the presence of the challenges faced.</div><div>A cross-section of the partners’ results is presented and briefly discussed, to provide an overview of the work done, and to encourage accessing and studying the project deliverables that are open to the public. Furthermore, the partners’ experiences made during the project have been evaluated and are presented as good practice recommendations.</div><div>The paper ends with conclusions on the level of readiness of UA in SA modelling, on the determination of governing uncertainties, and on the analysis of SAM actions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422984","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":"Numerical study on the thermal hydraulic characteristics of ERVC system","authors":"Yujian Huang ,&nbsp;Mingjun Wang ,&nbsp;Suizheng Qiu ,&nbsp;Kui Zhang ,&nbsp;Wenxi Tian ,&nbsp;Zhen Zhang","doi":"10.1016/j.anucene.2024.110972","DOIUrl":"10.1016/j.anucene.2024.110972","url":null,"abstract":"<div><div>Under the serious accident of core melting, in-vessel retention (IVR) can end the accident process inside the pressure vessel as an emergency strategy. The ERVC can remove the residual heat of the core.</div><div>The paper utilizes the method of enhanced heat transfer to improve CHF threshold. In this paper, according to investigating the types of fins, three fin structures (Longitudinal fin, Rectangular fin, Cylindrical fins) are selected and placed on the thermal insulation layer, which change the internal structure of the flow channel to play the role of turbulence disturbance. For the multi-phase flow model, a boiling model (RPI model) is used, considering the momentum exchange between the two phases, like drag force, virtual mass force, and wall lubrication force, as well as interphase mass transfer and heat transfer. The mathematical physical model is verified for the slicing experiment of ULPU, and the calculated result is compared with experimental physical values, as well as the error is within acceptable ranges, which are in good agreement. The calculations show that the CHF effect of cylindrical fin is better than rectangular fin and longitudinal fin, since that the turbulence intensity of around cylindrical fin is stronger than rectangular fin and longitudinal fin. For the same fin, when the fins spacing are smaller and fins height are between 40–60 mm, the cylindrical fin geometry has better cooling effect on the core and the residual heat removing. The enhanced heat transfer effect of cylindrical fins can be improved by 21 %.The numerical simulation calculation results can provide certain reference for engineering design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422986","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 hybrid model for annular linear induction pump","authors":"Sixian Zhu ,&nbsp;Jien Ma ,&nbsp;Keming Bi ,&nbsp;Xin Yan ,&nbsp;Youtong Fang","doi":"10.1016/j.anucene.2024.110931","DOIUrl":"10.1016/j.anucene.2024.110931","url":null,"abstract":"<div><div>Annular linear induction pump (ALIP) shows great potential in transport liquid metal in nuclear power system. A novel hybrid model for ALIP is proposed. The magnetic network is introduced to deal with iron core saturation and slot shape effects. Analysis model is introduced to deal with the axial magnetic flux. Correction factors are introduced to deal with the end effect. Furthermore, a small-scale ALIP was designed and manufactured. The results of proposed hybrid model, the finite element method (FEM) and the experiments are compared and show high consistence.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422923","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":"Flow distribution analysis of annular fuel with spacer grid","authors":"Sipeng Wang ,&nbsp;Qindong Zhang ,&nbsp;Mingwei Li ,&nbsp;Zhuoyao Li ,&nbsp;Bao-Wen Yang","doi":"10.1016/j.anucene.2024.110957","DOIUrl":"10.1016/j.anucene.2024.110957","url":null,"abstract":"<div><div>The annular fuel has internal and external channels, whose flow distribution is an important issue in fuel design. And compared with the traditional solid fuel rods, the gap between annular fuels is narrower, leading to smaller coolant flow rate and weaker cooling capacity at the gap. Therefore, the flow distribution of the annular fuel bundle with spacer grid is studied. And the narrow gap between the rods posed a challenge to the design of spacer grid, so a spacer grid with single vertebral canal is adopted in this study, with simple structure and convenient manufacture. As result shows, the spacer grid exhibits a significant impact on flow distribution of inner and outer channels, and it also makes the coolant redistribute in outer channel, improving local high temperature phenomenon at gaps. This study has reference significance for the flow distribution characteristics of spacer grid for annular fuel.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422919","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}