Annals of Nuclear Energy最新文献_第6页

Framework for the correct treatment of model input parameters for Bayesian updating problems in nuclear engineering 正确处理核工程中贝叶斯更新问题模型输入参数的框架

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-05 DOI: 10.1016/j.anucene.2024.110930

Michał Jędrzejczyk , Piotr Kopka , Basma Foad

{"title":"Framework for the correct treatment of model input parameters for Bayesian updating problems in nuclear engineering","authors":"Michał Jędrzejczyk , Piotr Kopka , Basma Foad","doi":"10.1016/j.anucene.2024.110930","DOIUrl":"10.1016/j.anucene.2024.110930","url":null,"abstract":"<div><div>Bayesian updating (BU) tools are increasingly used in nuclear engineering for inverse uncertainty quantification (IUQ) and calibration combined with uncertainty reduction. Their goals are quantifying or reducing the uncertainty of some of the uncertain model input parameters. Since it is often the case that available experimental data only allows for updating the most influential input parameters, researchers often ignore the less important ones during BU. This paper explores the consequences of neglecting the uncertainties of uncalibrated model input parameters (UMIP). It also proposes how to include them properly and which BU algorithms are the best choices for various types of inverse problems. The analysis is based on exploring two toy problems and one in nuclear engineering concerning multiplication factor calculations. The results clearly show that the improper treatment of UMIP during BU often leads to underestimating posterior uncertainties — either of the calibrated input parameters or the simulated integral parameters, depending on how the BU was conducted. The proposed methods of correct UMIP treatment will improve the rigorousness of the BU processes and boost confidence in the resulting posterior distributions.</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":"142422943","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}

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Californium-252 production at the High Flux Isotope Reactor − I: Validation study using campaign data 高通量同位素反应堆的锎-252 生产--I：利用运行数据进行验证研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-04 DOI: 10.1016/j.anucene.2024.110960

Donny Hartanto, David Chandler, Hailey Green, Jin Whan Bae, Kevin M. Burg, Yves Robert, Carol Sizemore

{"title":"Californium-252 production at the High Flux Isotope Reactor − I: Validation study using campaign data","authors":"Donny Hartanto, David Chandler, Hailey Green, Jin Whan Bae, Kevin M. Burg, Yves Robert, Carol Sizemore","doi":"10.1016/j.anucene.2024.110960","DOIUrl":"10.1016/j.anucene.2024.110960","url":null,"abstract":"<div><div>This paper presents a series of <sup>252</sup>Cf production validation and code-to-code comparison studies performed based on data from the production campaigns at the High Flux Isotope Reactor (HFIR). These studies support efforts to convert HFIR from using highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. HFIR must maintain its world-class performance and missions following this conversion, and because <sup>252</sup>Cf is a vital neutron-emitting radioisotope used for a variety of high-impact applications (e.g., reactor startup, cancer treatment), the ability to efficiently produce <sup>252</sup>Cf must be preserved. In this work, the HFIRCON, Shift, ORIGEN, and TCOMP codes were deployed, and several sets of data libraries were investigated to better understand the calculation codes and the data biases. As-loaded target composition data, as-run irradiation history data, and post-irradiation measurements from recent multi-cycle irradiation campaigns of the HEU core were used to validate and determine methodology biases. The findings demonstrated a good agreement, with results falling within 3 standard deviations of measurements. This paper lays the ground work for the second paper, which evaluates and compares <sup>252</sup>Cf production and safety metrics with the HEU core and a proposed LEU core.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422990","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

Advancing source reactor-type discrimination using machine learning techniques and SFCOMPO-2.0 experimental database 利用机器学习技术和 SFCOMPO-2.0 实验数据库推进源反应堆类型辨别工作

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-04 DOI: 10.1016/j.anucene.2024.110952

Tianxiang Wang, Hao Yang, Shengli Chen, Cenxi Yuan

{"title":"Advancing source reactor-type discrimination using machine learning techniques and SFCOMPO-2.0 experimental database","authors":"Tianxiang Wang, Hao Yang, Shengli Chen, Cenxi Yuan","doi":"10.1016/j.anucene.2024.110952","DOIUrl":"10.1016/j.anucene.2024.110952","url":null,"abstract":"<div><div>In recent years, nuclear forensic analysis has become crucial due to the growing global threat of nuclear terrorism and smuggling. Since 2005, extensive research has been conducted on identifying the origin of spent nuclear fuel, focusing on the source reactor-type discrimination, <sup>235</sup>U enrichment of the fresh fuel, and the fuel exposure in the reactor (known as burnup). However, the majority of research relies on computed databases, which may lead to tracing discrepancies compared with actual situations. The present study employs the isotopic measurements from the experimental SFCOMPO-2.0 database to predict nuclear reactor types using Factor Analysis (FA) and various machine learning classification algorithms. The results reveal that FA is an effective method for dimension reduction and visualization. The FA-KNN, Random Forest (RF), and Multilayer Perceptron (MLP) algorithms are applied using a consistent dataset partition to ensure unbiased comparisons. The prediction results based on 10-fold stratified cross-validation are quite promising and the Receiver Operating Characteristic (ROC) curves for multi-class classification confirm the excellent generalization ability of models. Therefore, the application of machine learning techniques is highly effective for reactor-type forensics analysis, especially for RF and MLP.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422989","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}

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A comparative analysis of detailed and reduced CFD approaches to model wire-wrapped fuel bundles for LMFBRs applications 详细和简化 CFD 方法的比较分析，以模拟低密度纤维束燃料元件应用中的线包燃料束

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-03 DOI: 10.1016/j.anucene.2024.110937

O. Halim , F. Galleni , N. Forgione , I. Di Piazza , A. Pucciarelli

引用次数: 0

Intergranular fracture induced by helium bubbles segregation in NiFe single-phase concentrated solid solution alloys 镍铁单相浓固溶合金中氦气泡偏析诱发的晶间断裂

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-03 DOI: 10.1016/j.anucene.2024.110959

Jie Li , Xinhua Yang , Peng Wang , Qin Qian

{"title":"Intergranular fracture induced by helium bubbles segregation in NiFe single-phase concentrated solid solution alloys","authors":"Jie Li , Xinhua Yang , Peng Wang , Qin Qian","doi":"10.1016/j.anucene.2024.110959","DOIUrl":"10.1016/j.anucene.2024.110959","url":null,"abstract":"<div><div>Single-phase concentrated solid solution alloys (SP-CSAs) are potential nuclear structural materials with excellent irradiation resistance. Under irradiation environment, helium (He) atoms are easily segregated and stored at grain boundaries (GBs), thus causing He embrittlement of materials. The intergranular fracture behavior of NiFe SP-CSAs caused by He bubbles segregation is studied with molecular dynamics (MD) method. The effects of Fe atom concentration, GB type and He atom number in He bubbles on the intergranular fracture are analyzed. Since the disordered atoms impede crack initiation and propagation, NiFe SP-CSAs exhibit stronger fracture resistance than pure Ni. With the increase of strain, the number of disordered atoms at the GB increases further, so that the fracture strain of NiFe SP-CSAs with symmetric tilt GB decreases. With the increase of Fe atom concentration, Fe atoms inhibit the intergranular crack initiation and propagation in NiFe SP-CSAs more powerfully. In addition, the increase of the He atom number in the He bubble promotes the fracture. The formation of SP-CSAs is also beneficial to enhancing intergranular fracture resistance of alloys with twist GBs and asymmetrically tilt GBs, and the impact on the latter is more pronounced.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422922","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

Fuel performance modelling of Cr-coated Zircaloy cladding under DBA/LOCA conditions DBA/LOCA 条件下铬涂层锆合金包层的燃料性能建模

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-03 DOI: 10.1016/j.anucene.2024.110950

Pau Aragón , Francisco Feria , Luis E. Herranz , Arndt Schubert , Paul Van Uffelen

{"title":"Fuel performance modelling of Cr-coated Zircaloy cladding under DBA/LOCA conditions","authors":"Pau Aragón , Francisco Feria , Luis E. Herranz , Arndt Schubert , Paul Van Uffelen","doi":"10.1016/j.anucene.2024.110950","DOIUrl":"10.1016/j.anucene.2024.110950","url":null,"abstract":"<div><div>Chromium coatings are being developed for advanced technology fuel (ATF) claddings, offering negligible corrosion during normal operation, improved resistance to high-temperature steam oxidation, and superior high-temperature strength, the latter two being of utmost relevance during design basis accidents (DBAs). Demonstrating the improved response of Cr-coated Zircaloy requires the development or extension of fuel performance codes to coating simulations.</div><div>In this work, material models and correlations for Cr-coated Zircaloy cladding have been derived or obtained from the literature and implemented into TRANSURANUS and the FRAPTRAN-TUmech suite. These extended tools have been used to simulate two complementary LOCA tests: QUENCH-L1 rod 4 (out-of-pile bundle test on fresh Zircaloy cladding) and IFA-650.10 (in-pile single rod test on high-burnup Zircaloy-UO<sub>2</sub> fuel), enabling a gradual cross-verification of results between codes and a comparative performance analysis between coated and uncoated cladding.</div><div>The results indicate negligible impact of coating properties other than creep on the burst time. While the superior high-temperature creep resistance of coated cladding slightly delays the burst time, additional burst data would be necessary to draw sound conclusions on the balloon size. Regarding the modelling approach, treating the coated cladding as a composite material through the definition of effective properties might result in worse performance relative to uncoated cladding, contradicting experimental observations. Therefore, the separate modelling of the coating and the cladding is recommended.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422988","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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On the convergence of the fixed point method for solving neutron transport alpha eigenvalue problems 论解决中子输运阿尔法特征值问题的定点法的收敛性

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-02 DOI: 10.1016/j.anucene.2024.110620

Britton Chang

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Flow-heat coupling analysis of the 1/4 symmetrical CAP1400 nuclear island loop based on the source term approach 基于源项方法的 1/4 对称 CAP1400 核岛环路流热耦合分析

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-01 DOI: 10.1016/j.anucene.2024.110926

Yeming Lu , Zhenyang Guo , Meina Zhang , Meina Zhang , Xiaomo Jiang , Xiaofang Wang

{"title":"Flow-heat coupling analysis of the 1/4 symmetrical CAP1400 nuclear island loop based on the source term approach","authors":"Yeming Lu , Zhenyang Guo , Meina Zhang , Meina Zhang , Xiaomo Jiang , Xiaofang Wang","doi":"10.1016/j.anucene.2024.110926","DOIUrl":"10.1016/j.anucene.2024.110926","url":null,"abstract":"<div><div>For the large-scale CAP1400 nuclear island, characterized by its numerous and extensive equipment, each component plays distinct roles and interacts with others, posing significant challenges for numerical simulations of the entire island loop. Traditional studies of nuclear coolant pumps often focus on the pump in isolation, which inadequately captures its complete operational characteristics within the integrated system. To efficiently and accurately determine the operational characteristics of the nuclear coolant pump within the complete nuclear island system, this study simplifies the models of the evaporator and reactor by employing a porous medium approach. The accuracy is ensured through the inclusion of momentum and heat source terms. The models of the evaporator, reactor, and main nuclear pump are interconnected using the momentum source term pipeline, and a simplified 1/4 symmetrical nuclear island loop simulation platform is developed, taking advantage of the mirror symmetry in the island’s circuit layout. Via comparative analysis, it can be found that: 1) The simulation model was validated using CAP1400 operational and experimental data. Temperature calculation errors at key sections were below 1.5%, and the numerical errors for the pump head and efficiency were within 0.55% and 1.87%, respectively, demonstrating high accuracy and reliability. 2) Integrating the isolated nuclear coolant pump into the closed-loop system resulted in an increase in head by 0.21 m and a decrease in efficiency by 2.64%, indicating significant energy loss within the pump’s internal channels and the impact of the closed-loop system on pump performance. 3) Transient simulation revealed that the closed-loop system disrupted the pump’s symmetric balance, leading to uneven pressure and temperature distribution and higher pressure pulsation coefficients at various monitoring points, which could affect system stability and safety.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422920","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}

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Dynamic simulation of core power control for SPWR based on dynamic model bank 基于动态模型库的 SPWR 核心功率控制动态模拟

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-30 DOI: 10.1016/j.anucene.2024.110944

Zheng Li, Xiaoyu Li, Linna Wang, Wenjie Zeng

引用次数: 0

Development and verification of an MC/MOC two-step scheme for neutronic analysis of FCM-fueled micro gas-cooled reactor 开发并验证用于 FCM 燃料微型气冷堆中子分析的 MC/MOC 两步方案

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-09-29 DOI: 10.1016/j.anucene.2024.110940

Kuaiyuan Feng , Qufei Song , Yuyang Shen , Lei Lou , Yao Xiao , Hui Guo , Hanyang Gu

{"title":"Development and verification of an MC/MOC two-step scheme for neutronic analysis of FCM-fueled micro gas-cooled reactor","authors":"Kuaiyuan Feng , Qufei Song , Yuyang Shen , Lei Lou , Yao Xiao , Hui Guo , Hanyang Gu","doi":"10.1016/j.anucene.2024.110940","DOIUrl":"10.1016/j.anucene.2024.110940","url":null,"abstract":"<div><div>Gas-cooled microreactors are known for compact designs, high thermal-to-electric efficiencies, long refueling cycles, and flexible deployment capabilities, representing a groundbreaking solution to address the energy requisites of special scenarios. Fully ceramic microencapsulated (FCM) fuel is widely used in gas-cooled microreactors, bringing challenges to neutronic analysis methods. In this paper, an Monte Carlo/Method of Characteristics (MC/MOC) two-step scheme is developed and verified based on the reference case. In this scheme, the continuous-energy Monte Carlo calculations are used for reference calculation and multi-group cross-section generation. The multi-group Monte Carlo calculations are used for multi-group cross-section verification and the MOC solver verification. Fuel multi-group cross-sections are generated with the explicit fuel assembly model by continuous-energy Monte Carlo calculations, and structure multi-group cross-sections are generated with the simplified core model by continuous-energy Monte Carlo calculations. The core calculations are conducted with the MOC calculations. For verification, parameters such as power distribution, neutron spectrum, and control devices worth will be compared. Core calculation results show that the relative errors of MOC results are within −329 pcm, ±5%, ±6%, and ± 2 % for K<sub>eff</sub>, power distribution, neutron spectrum, and control device worth, separately. Moreover, the computation cost of MOC is only 6.6 % of the reference computation cost. The figure of merit results show that the MC-MOC scheme exhibits improved computational efficiency for neutronic analysis of FCM-fueled micro gas-cooled reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142357881","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