Annals of Nuclear Energy最新文献_第5页

Research on fluid flow and heat transfer characteristics in a three-dimensional condenser 三维冷凝器中的流体流动和传热特性研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-09 DOI: 10.1016/j.anucene.2024.110967

Zhiqiang Duan, Yuan Tian, Siyuan Wang, Ling Long, Jianjun Deng

{"title":"Research on fluid flow and heat transfer characteristics in a three-dimensional condenser","authors":"Zhiqiang Duan, Yuan Tian, Siyuan Wang, Ling Long, Jianjun Deng","doi":"10.1016/j.anucene.2024.110967","DOIUrl":"10.1016/j.anucene.2024.110967","url":null,"abstract":"<div><div>The condenser plays a crucial role in nuclear power plants, impacting equipment economics and safety through shell-side two-phase flow and heat transfer. However, existing research has oversimplified the tube bundles, lacking comprehensive information for optimal performance. In this research, a three-dimensional method incorporating mixture and multiphase flow condensation models was used to investigate flow behavior and heat transfer characteristics without simplifying the internal structure. Calculation details were enhanced by omitting the porous medium approach. The numerical model achieved reasonable accuracy when compared to theoretical calculations for a range of steam mass flow rates. Analysis of numerical results, including pressure, velocity, temperature, air mass fraction, and heat transfer coefficient, revealed that steam flow rate and air mass fraction were key factors influencing heat transfer. This research demonstrates the method’s capability to capture intricate calculation details, providing valuable insights for optimization design considerations in condenser performance.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422949","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

High-fidelity multiphysics modeling of pulsed reactor heat generation in the Annular Core Research Reactor fuel using Serpent 2 利用 Serpent 2 对环形堆芯研究堆燃料中的脉冲反应堆发热进行高保真多物理场建模

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-09 DOI: 10.1016/j.anucene.2024.110954

Emory Colvin, Todd S. Palmer

{"title":"High-fidelity multiphysics modeling of pulsed reactor heat generation in the Annular Core Research Reactor fuel using Serpent 2","authors":"Emory Colvin, Todd S. Palmer","doi":"10.1016/j.anucene.2024.110954","DOIUrl":"10.1016/j.anucene.2024.110954","url":null,"abstract":"<div><div>Sandia National Laboratories’ Annular Core Research Reactor is a unique pulsed reactor using UO<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>-BeO fuel. Unlike TRIGA reactors, which often operate in steady-state with infrequent pulsing, the ACRR operates almost exclusively in pulsed mode. Throughout the last 15 years, computational reactor physics analyses of the ACRR have involved either kinetic simulations with reduced physics approaches to the neutron distribution or a detailed Monte Carlo criticality simulation in steady-state. This paper presents an effort to perform time-dependent Monte Carlo modeling of pulsed operation of the ACRR for the purpose of determining the heat generation in the fuel. This time and space-dependent volumetric heat generation will serve as a source term for future analysis of the physical characteristics of the fuel after many pulses to inform decisions about ACRR operation and the future design of ACRR-like reactors. The Serpent Monte Carlo code is coupled to a simple Python script providing updated fuel temperatures after each time step, allowing on-the-fly adjustment of fuel cross sections. Beginning with a simple pin of UO<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>-BeO fuel, models increase in complexity to the full model of the ACRR. Results from the full model of the ACRR are compared to experimental results, and computational efficiency and heat generation plots are discussed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422950","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

Monte Carlo based absolute efficiency calibration of power reactor spent fuel NDA measurements 基于蒙特卡洛的动力堆乏燃料 NDA 测量绝对效率校准

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-08 DOI: 10.1016/j.anucene.2024.110953

Péter Kirchknopf , Zoltán Kató , Csongor Kristóf Szarvas , Péter Völgyesi , Imre Szalóki

{"title":"Monte Carlo based absolute efficiency calibration of power reactor spent fuel NDA measurements","authors":"Péter Kirchknopf , Zoltán Kató , Csongor Kristóf Szarvas , Péter Völgyesi , Imre Szalóki","doi":"10.1016/j.anucene.2024.110953","DOIUrl":"10.1016/j.anucene.2024.110953","url":null,"abstract":"<div><div>Experiments have been carried out at Paks Nuclear Power Plant for the purpose of spent fuel burnup characterization using gamma-ray spectrometry. Obtaining absolute quantitative information, e.g. fission product activities, from the spectra requires accurate knowledge of the detection efficiency. Due to nature of the measurement conditions, experimental calibration was unfeasible, and the Monte Carlo particle transport method was selected to calculate the efficiency. The model building process is presented, which involves X-ray radiography of the germanium detector, optimization of the dead layer on the outside of the crystal, and a solution to the challenging low-efficiency simulation problem. The simulations were validated using measurements carried out at the Paks reactor units together with verified experimental data from the SFCOMPO library. The constructed model proved to be accurate to the results determined by empirical methods within ±3 % error, promising to be a reliable basis for future applications that require spent fuel characterization.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422947","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}

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The water ingress analysis on steam generator heat-exchange tube rupture accident of high temperature gas-cooled reactor 高温气冷堆蒸汽发生器换热管破裂事故的进水分析

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-08 DOI: 10.1016/j.anucene.2024.110968

Hongming Huang , Rongshun Xie , Songsong Liu , Xiaotong Wu , Guotai Cheng , Yaoli Zhang

引用次数: 0

Nuclear data uncertainty propagation in continuous-energy Monte Carlo calculations 连续能蒙特卡罗计算中的核数据不确定性传播

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-08 DOI: 10.1016/j.anucene.2024.110955

Alexander Aures, Thomas Eisenstecken, Ekaterina Elts, Robert Kilger

{"title":"Nuclear data uncertainty propagation in continuous-energy Monte Carlo calculations","authors":"Alexander Aures, Thomas Eisenstecken, Ekaterina Elts, Robert Kilger","doi":"10.1016/j.anucene.2024.110955","DOIUrl":"10.1016/j.anucene.2024.110955","url":null,"abstract":"<div><div>The XSUSA method is a well-established stochastic sampling method for propagating nuclear data uncertainties through multigroup neutron transport calculations. To benefit from the advantages of Monte Carlo transport codes, namely modeling complex geometries and using continuous-energy nuclear data, an extension to XSUSA is proposed which allows perturbing continuous-energy nuclear data using multigroup nuclear data covariances. To verify the extension, sensitivity profiles of nuclear reactions are calculated via direct perturbation for the benchmark problems Jezebel, Godiva, LEU-SOL-THERM-002. The sensitivity profiles agree well with those obtained from TSUNAMI and Serpent. Secondly, the extension to XSUSA is applied to produce randomly sampled continuous-energy data libraries using the covariance libraries of SCALE 6.2. With these data libraries, samples of Serpent calculations are performed for Jezebel, Godiva, LEU-SOL-THERM-002, and the TMI-1 pin cell of the OECD/NEA LWR-UAM benchmark. For each problem, the multiplication factor uncertainty agrees well with the one from TSUNAMI.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422948","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 a computational scheme based on the DRAGON5 code for the neutronic study of VVER-type reactor rods and assemblies 开发基于 DRAGON5 代码的计算方案，用于 VVER 型反应堆棒材和组件的中子研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-07 DOI: 10.1016/j.anucene.2024.110961

Cyprien Richard , Mathias François , Lucas Fede , Alain Hébert

{"title":"Development of a computational scheme based on the DRAGON5 code for the neutronic study of VVER-type reactor rods and assemblies","authors":"Cyprien Richard , Mathias François , Lucas Fede , Alain Hébert","doi":"10.1016/j.anucene.2024.110961","DOIUrl":"10.1016/j.anucene.2024.110961","url":null,"abstract":"<div><div>Open source modeling of VVER-type reactors could become a medium-term objective in Eastern Europe. As the deterministic code DRAGON5 could meet such a need, we confronted DRAGON5 against a stochastic reference code, SERPENT2. Our validation comprises 7 cells and 4 assemblies from the Khmelnitsky-2 reactor in Ukraine, within a wide range of heterogeneity levels in fuel composition. Two calculation schemes have been developed and compared. The first, the ALAMOS scheme, is highly discretized in energy and spatial resolution, while the second, the REL2005-like scheme, is calculated in two levels (one highly discretized in energy and the other highly discretized in space). In the majority of cases studied, both schemes offer satisfactory accuracy (e.g. less than 300 pcm in <span><math><msub><mrow><mi>k</mi></mrow><mrow><mtext>eff</mtext></mrow></msub></math></span>), although there are difficulties related to energy deposition with gadolinium-poisoned fuel. While showing significantly poorer results than the ALAMOS scheme, the REL2005-like scheme offers lower computation times and major avenues for improvement remain to be explored. This work offers a first step towards the simulation of VVER-type reactors in DRAGON5, and paves the way for full-core simulations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422987","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

Analysis of swarm flow and bubble residence time under pool scrubbing conditions 水池洗涤条件下的群流和气泡停留时间分析

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-07 DOI: 10.1016/j.anucene.2024.110956

Fanli Kong, Xu Cheng

{"title":"Analysis of swarm flow and bubble residence time under pool scrubbing conditions","authors":"Fanli Kong, Xu Cheng","doi":"10.1016/j.anucene.2024.110956","DOIUrl":"10.1016/j.anucene.2024.110956","url":null,"abstract":"<div><div>In severe accidents of nuclear power plants, large amounts of fission products existing as radioactive aerosols are released. Pool scrubbing plays an important role in the removal of radioactive aerosols. Bubble residence time is one of the key parameters to determine the efficiency of aerosol removal, especially in the swarm flow region which makes a very important contribution to the total aerosol removal. In this study, the Euler-Euler-Lagrangian approach is built to track the evolution of bubble motion and to determine the bubble residence time in the liquid pool. Specifically, the Euler-Euler two-fluid approach is utilized to resolve the flow field of gas and liquid phases, while the Lagrangian approach is employed to track the discrete bubbles and to obtain the bubble residence time. The results reveal that the present approach is feasible to predict the bubble dynamics and residence time in the liquid pool. Bubble residence time is dependent on the initial position, where bubbles deviating from the central region could remain inside the liquid pool for a longer physical time. The bubble diameter, volume flow rate and submergence height are key parameters affecting the bubble residence time. And comparison between the simulated bubble residence time and the model-predicted results is carried out, indicating the discrepancy of simulated residence time and limitations of the existing model at high volume flow rate and high submergence.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422945","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

Alternative core configurations analysis to improve the neutronics performance of modular gas cooled fast reactor 提高模块式气冷快堆中子性能的备选堆芯配置分析

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-07 DOI: 10.1016/j.anucene.2024.110951

Shohanul Islam, Md Tanvir Ahmed

{"title":"Alternative core configurations analysis to improve the neutronics performance of modular gas cooled fast reactor","authors":"Shohanul Islam, Md Tanvir Ahmed","doi":"10.1016/j.anucene.2024.110951","DOIUrl":"10.1016/j.anucene.2024.110951","url":null,"abstract":"<div><div>This study investigates the neutronics characterization of the Allegro-75 MW<sub>th</sub> modular reactor by analyzing three alternative heterogeneous core configurations-axial, radial, axial + radial across three fuel candidates-UPuC, UPuN, and UPuO along with reflector materials namely ZrC, SiC, BeO, and Zr<sub>3</sub>Sc<sub>2</sub> to improve neutronics performance, identify the most suitable core configuration and optimal axial reflector thickness. The study revealed that axial + radial heterogeneous core configuration exhibited better performance across each fuel type compared to other heterogeneous models. UPuC with axial + radial heterogeneity was identified as the optimal model as it demonstrated cycle length over ten years, satisfactory neutron spectrum, uniform neutron flux distribution, low radial and axial PPF, high beta effective, and negative Doppler Constant. Analyzing reflector materials with the most suitable fuel model revealed that the optimum axial reflector thickness is 60 cm for all reflector models where BeO emerged as the most favorable reflector due to its superior results in other neutronics parameters.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422944","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 study on in-situ characterization technology development for clearance verification of radioactive waste from nuclear decommissioning 核退役放射性废物清除核查原位表征技术开发研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-05 DOI: 10.1016/j.anucene.2024.110945

Kyungmin Kim, Minseung Ko, Sangtae Kim, Yongsoo Kim

{"title":"A study on in-situ characterization technology development for clearance verification of radioactive waste from nuclear decommissioning","authors":"Kyungmin Kim, Minseung Ko, Sangtae Kim, Yongsoo Kim","doi":"10.1016/j.anucene.2024.110945","DOIUrl":"10.1016/j.anucene.2024.110945","url":null,"abstract":"<div><div>Although the clearance level of every radioactive nuclide was published by the IAEA to promote the recycling and reuse of decontaminated radioactive waste worldwide, technical and regulatory issues have always been raised around the application of the criteria. Therefore, several countries are developing in-situ characterization equipment or apparatus for on-site verification to check if the clearance criteria is met.</div><div>In this study authors developed a pilot radiation detection and measurement system using in-situ characterization technology to solve the issues, which consists of a 3D scanning camera system and a built-in Monte Carlo simulation program. Measurement results show that MDA (Minimum Detectable Activity) of the current design was <span><span>indisputably</span><svg><path></path></svg></span> below the clearance level and built-in Monte Carlo simulation package closely predicts the measurements results with the error of less than 5%. This implicates that it can determine with enough margin whether the radioactivity level of decontaminated metallic components meets the clearance criteria at decommissioning site or not.</div><div>Practically when we measure the radioactivity from gamma ray source mass attenuation always takes place during the photon transports through the medium. In fact, the reduction depends on the material, shapes, and radioactive sources. In this study the reduction factors were experimentally examined according to the influencing parameters and the results were saved as DCF (Density Correction Factor) in the data base. As expected, it turned out that the factor is somewhat affected by medium material and radioactive sources, but it is basically proportional to the distance of gamma ray passage.</div><div>It is expected that upgraded design with more accurate and reliable instruments can make it easier for regulators to accept the application of the in-situ characterization technology on-site.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422991","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

Radiation shielding properties of sustainable concrete with novel plastering techniques 采用新型抹灰技术的可持续混凝土的辐射屏蔽性能

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-05 DOI: 10.1016/j.anucene.2024.110958

Mohamed A.E.M. Ali , Mohammed A.Y. Hafez , Nabil M. Nagy , Neveen S. Abed

{"title":"Radiation shielding properties of sustainable concrete with novel plastering techniques","authors":"Mohamed A.E.M. Ali , Mohammed A.Y. Hafez , Nabil M. Nagy , Neveen S. Abed","doi":"10.1016/j.anucene.2024.110958","DOIUrl":"10.1016/j.anucene.2024.110958","url":null,"abstract":"<div><div>In concrete applications. Major/critical applications of such concrete are radiation-shielding facilities. Both steel slag and silica fume are examples of common by-product materials that can be used as a replacer of aggregates and cement. Thus, in this research work, steel slag was utilized as heavy aggregate in concrete production besides silica fume to present sustainable concrete mixtures probably with better radiation-shielding properties. Different cementitious plasters were applied on the conducted sustainable concrete mixture using different powdery materials; hematite, magnetite, barite, bentonite, and steel slag powders in addition to nano-titanium dioxide as full replacers for sand. The proposed plasters were presented to determine the optimum plaster technique in terms of static performance and attenuation capability against gamma and neutron radiations. The results exhibited that utilizing steel slag and silica fume in concrete mixtures enhanced compressive strength by up to 9.09 % compared to conventional concrete, while the addition of nano-titanium to conventional plaster led to superior enhancement in the compressive strength by up to 38.65 % relative to traditional plaster. Conversely, fully replacing conventional silica sand with the abovementioned powdery materials generally reduced the compressive strength of cementitious plasters by up to 30.83 %. However, the radiation shielding properties against Cs-137, and Co-60 energies have been enhanced by up to 20 % and 26 %, respectively.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422942","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