Nuclear Engineering and Design最新文献

Material cross-sections real-time inversion from neutron detector data for nuclear reactor digital twin 核反应堆数字孪生体中子探测器数据的材料截面实时反演

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-07-01 DOI: 10.1016/j.nucengdes.2025.114296

Honghang Chi, Jiancheng Chen, Yahui Wang, Yu Ma

{"title":"Material cross-sections real-time inversion from neutron detector data for nuclear reactor digital twin","authors":"Honghang Chi, Jiancheng Chen, Yahui Wang, Yu Ma","doi":"10.1016/j.nucengdes.2025.114296","DOIUrl":"10.1016/j.nucengdes.2025.114296","url":null,"abstract":"<div><div>The Nuclear Reactor Digital Twin (NRDT) has garnered significant attention in recent years. One of the crucial aspects in NRDT is the real-time inversion of nuclear reactor core material cross-sections during reactor operation. In general, an inverse problem is solved by combining multiple iterations of the forward problem with an optimization algorithm. Even though the development of a surrogate model has significantly enhanced the computational efficiency of forward problems, the iteration process still poses a challenge to real-time inversion. To address this problem, this paper presents a real-time inverse problem solver (RIPS). During the offline stage, RIPS establishes a mapping between the sparse neutron detector data and the cross-sections through the reduced-order model and radial basis function. During the online stage, the corresponding cross-section can be calculated directly using the mapping and neutron detector data. Since the RIPS eliminates the multiple iterations of traditional methods, the efficiency of RIPS can be improved by orders of magnitude, and enables real-time online calculation. Three typical numerical benchmarks are tested for verification in this paper, which proves that the maximum relative error of RIPS does not exceed 0.54 % and the average relative error does not exceed 0.1986 %. Furthermore, for each test case, the calculation time of RIPS is within 0.01 s. This work can provide useful suggestions and applications, and further development in cross-section inversion.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114296"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144517698","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

Review of conceptual design and fundamental research related to the passive residual heat removal system in molten salt reactors 熔盐堆被动余热排出系统的概念设计与基础研究综述

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-07-01 DOI: 10.1016/j.nucengdes.2025.114275

Shuaiyu Xue , Chong Zhou , Pinyan Huang , Yang Zou

{"title":"Review of conceptual design and fundamental research related to the passive residual heat removal system in molten salt reactors","authors":"Shuaiyu Xue , Chong Zhou , Pinyan Huang , Yang Zou","doi":"10.1016/j.nucengdes.2025.114275","DOIUrl":"10.1016/j.nucengdes.2025.114275","url":null,"abstract":"<div><div>The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes such as inherent safety, sustainable development, nuclear non-proliferation, natural resource protection, and economic efficiency. Passive residual heat removal systems for MSRs predominantly encompass Direct Reactor Auxiliary Cooling Systems (DRACS), Salt Discharge Tanks Residual Heat Removal Systems (DTRHRS), and Heat Pipes Residual Heat Removal Systems (HPRHRS). This study introduces an innovative Secondary Side Passive Residual Heat Removal System (SSHRS) for MSRs. The SSHRS employs the primary heat exchanger to dissipate the residual heat from the fuel salt in the primary loop, eliminating the necessity for an additional residual heat removal exchanger and enhancing economic efficiency. The SSHRS approach prevents direct heat transfer from the fuel salt to the environment, mitigates the risk of radioactive material leakage, and bolsters safety. Furthermore, this study also made a horizontal comparison of the advantages and disadvantages of DRACS, DTRHRS, and SSHRS in terms of safety and economy, and discussed the future research directions of passive residual heat removal from molten salt reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114275"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144517685","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 feasibility of dynamic rod worth measurement using fission chamber detector signal in HPR1000 利用HPR1000裂变室探测器信号动态测量棒值的可行性研究

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-07-01 DOI: 10.1016/j.nucengdes.2025.114255

Jun Lin , Xianghui Lu , Changyou Zhao , Daifu Wang , Qingyu Gao , Yan Zhao , Xianjun Li , Mingtao He

{"title":"Study on the feasibility of dynamic rod worth measurement using fission chamber detector signal in HPR1000","authors":"Jun Lin , Xianghui Lu , Changyou Zhao , Daifu Wang , Qingyu Gao , Yan Zhao , Xianjun Li , Mingtao He","doi":"10.1016/j.nucengdes.2025.114255","DOIUrl":"10.1016/j.nucengdes.2025.114255","url":null,"abstract":"<div><div>The control rod worth measurement is one of the most important items during the reactor physical startup tests. Currently the dynamic rod worth measurement (DRWM) based on the signal from non-gamma compensated ion chamber is generally applied on worldwide nuclear power plants, which results in the temporary unavailability of one power range detector and possible insufficiency of zero power physical test range. The fission chamber detector with broader measuring range can transmit effective signal to the digital control system and makes it possible to be applied on the DRWM technique. The calculation model of the HPR1000 ex-core fission chamber detector signal and also the corresponding DRWM spatial-correction factors are built based on OpenMC and PCM and the verification of the calculation model is launched using the measured data from the first cycle of FANGCHENGGANG Unit 3 and Unit 4. The result shows that the calculated and measured variation of current during insertion of control rod are consistent and the DRWM control rod worth using fission chamber detector signals agrees well with that using power range detector signals as well as the theoretical prediction, which proves the feasibility to apply fission chamber detector signals in DRWM.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114255"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144523903","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

Research on severe accident diagnosis method based on PCA and DT for a small modular PWR 基于PCA和DT的小型模块化压水堆严重事故诊断方法研究

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-07-01 DOI: 10.1016/j.nucengdes.2025.114285

Xiaolong Bi, Peiwei Sun, Xinyu Wei

{"title":"Research on severe accident diagnosis method based on PCA and DT for a small modular PWR","authors":"Xiaolong Bi, Peiwei Sun, Xinyu Wei","doi":"10.1016/j.nucengdes.2025.114285","DOIUrl":"10.1016/j.nucengdes.2025.114285","url":null,"abstract":"<div><div>Small modular pressurized water reactor (SMPWR) is a new trend in the development of nuclear energy today. SMPWRs usually adopt the integrated arrangement and use many passive safety systems, which have high inherent safety. However, despite all precautions, the severe accident (SA) cannot be completely avoided. It is necessary to conduct rapid and timely diagnosis for these SAs. Different from the traditional SA diagnosis methods based on signals and knowledge, the data-based method is adopted to study SA diagnosis of SMPWR in this study. First, the SA state monitoring method based on principal component analysis (PCA) is proposed to realize rapid and accurate distinction between steady-state and abnormal conditions. Then, the SA classification diagnosis method based on decision tree (DT) is proposed to realize accurate diagnosis of initial events and whether the entry conditions of SA management guideline (SAMG) are met. The key parameters selected through feature selection can provide supplement and reference for the selection of monitoring parameters in SAMG and the determination of the instrument list for instrument availability analysis under SA conditions. The fault diagnosis method proposed in this paper can provide the reference and basis for the SA diagnosis and the design of the operator’s SA handling auxiliary system in the SMPWR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114285"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144517686","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

Plant dynamics analysis of sodium-cooled fast reactors using Flownex code 使用Flownex代码对钠冷快堆进行电厂动力学分析

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-30 DOI: 10.1016/j.nucengdes.2025.114267

Vikram Govindarajan , Rajendrakumar M. , Suresh Kumar R. , Natesan K.

{"title":"Plant dynamics analysis of sodium-cooled fast reactors using Flownex code","authors":"Vikram Govindarajan , Rajendrakumar M. , Suresh Kumar R. , Natesan K.","doi":"10.1016/j.nucengdes.2025.114267","DOIUrl":"10.1016/j.nucengdes.2025.114267","url":null,"abstract":"<div><div>Plant dynamics analysis plays an important role in the design and operation of sodium-cooled fast reactors (SFRs). This paper presents the development and validation of a plant dynamic model for an SFR using the Flownex code, a general-purpose thermal-fluid simulation software. A general modeling philosophy is provided for building Flownex models to simulate key components of the SFR, including the core, plenum, pipelines, intermediate heat exchanger (IHX), and pump. A new user-defined script for SFR kinetics calculations has been developed, which enables neutron kinetics calculations based on the point kinetics model with reactivity feedback effects. The procedures for simulating heat sources, sinks, and the inclusion of temperature-dependent sodium fluid properties are discussed in detail. These models and approaches are designed to optimize execution speed while maintaining good accuracy based on practical experience with the code.</div><div>A plant dynamics model is developed for the “Fast Flux Test Facility (FFTF)” reactor using Flownex and is used to simulate the “Loss Of Flow Without Scram (LOFWOS)” test #13. The good agreement observed between the simulation results of various SFR parameters and the experimental data demonstrates the suitability of the Flownex code for advanced plant dynamics studies of SFRs. However, the current system modeling approach has certain limitations, primarily due to the neglect of spatial (multidimensional) effects and the simplified treatment of feedback reactivity components. Potential directions for future improvements are also discussed in this paper.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114267"},"PeriodicalIF":1.9,"publicationDate":"2025-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144513791","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 simulation of forced convection condensation from steam-air mixtures on vertical flat plates and evaluation of wall function 垂直平板上蒸汽-空气混合物强制对流冷凝的数值模拟及壁面功能的评价

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-30 DOI: 10.1016/j.nucengdes.2025.114289

Michio Murase , Yoichi Utanohara , Toshiya Takaki

{"title":"Numerical simulation of forced convection condensation from steam-air mixtures on vertical flat plates and evaluation of wall function","authors":"Michio Murase , Yoichi Utanohara , Toshiya Takaki","doi":"10.1016/j.nucengdes.2025.114289","DOIUrl":"10.1016/j.nucengdes.2025.114289","url":null,"abstract":"<div><div>We carried out numerical simulation of wall condensation from steam-air mixtures in forced convection (FC) flows on vertical flat plates by using the computational fluid dynamics (CFD) code FLUENT and the computational domain which simulated the flow channel in the SETCOM facility. We evaluated three items: the local Sherwood number <em>Sh<sub>y</sub></em>; the condensation heat flux <em>q<sub>c,y</sub></em>, and the wall function, which are used in CFD analysis with coarse computation cells. The computed mixture temperature <em>T<sub>g,cal</sub></em> was lower than the existing data <em>T<sub>g,</sub></em><sub>exp</sub> in the turbulent region, but the computed heat fluxes <em>q<sub>CFD</sub></em> along the flow direction in FC flows agreed well with the heat fluxes <em>q<sub>exp</sub></em> measured with the COPAIN and CONAN facilities (the standard deviation <em>s</em> of <em>q<sub>CFD</sub></em> to <em>q<sub>exp</sub></em> was <em>s</em> = 13 %). From the CFD results, we obtained the improved correlation for <em>Sh<sub>y</sub></em> to predict <em>q<sub>c,y</sub></em> for the viscous sublayer, the transition region, and the turbulent region, and we found the <em>q<sub>c,y</sub></em>/<em>q<sub>c,CFD</sub></em> values with the <em>Sh<sub>y</sub></em> correlation were within 1.0 ± 0.12 for the inlet mixture velocity of <em>u<sub>in</sub></em> = 5.47 m/s and the flow direction of <em>x</em> = 1–5 m. A logarithmic function for the modified dimensionless steam mass fraction <em>Y<sub>s,mod</sub></em><sup>+</sup> was proposed, and uncertainty of the values with the <em>Y<sub>s,mod</sub></em><sup>+</sup> function compared to the CFD results was ± 2 <em>s</em> = ±9.3 % for <em>u<sub>in</sub></em> = 5.47 m/s and <em>x</em> = 1–5 m. Finally, we discussed effects of <em>u<sub>in</sub></em> on <em>Sh<sub>y</sub></em> and <em>Y<sub>s,mod</sub></em><sup>+</sup>.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114289"},"PeriodicalIF":1.9,"publicationDate":"2025-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144517684","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

Solidification of Cs-bearing melting fly ash by cement and metakaolin-based geopolymers 水泥和偏高岭土聚合物固化含碳熔融粉煤灰

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-30 DOI: 10.1016/j.nucengdes.2025.114253

Kanako Toda , Lang Lang , Takumi Saito , Kazuto Endo , Kazuo Yamada , Yasumasa Tojo

{"title":"Solidification of Cs-bearing melting fly ash by cement and metakaolin-based geopolymers","authors":"Kanako Toda , Lang Lang , Takumi Saito , Kazuto Endo , Kazuo Yamada , Yasumasa Tojo","doi":"10.1016/j.nucengdes.2025.114253","DOIUrl":"10.1016/j.nucengdes.2025.114253","url":null,"abstract":"<div><div>The melting fly ash (MFA) stored in the Interim Storage Facility in Fukushima, Japan, contains concentrated Cs, requiring stabilization for final disposal. This study investigates the physical properties to evaluate the feasibility of the MFA conditioning with B-type Portland blast furnace slag (BFS) cement and Na/K-activated metakaolin-based geopolymers (NaGP and KGP), with the evaluation of the Cs leaching behavior from solidified forms. Because of the difference in flowability, the volume of the solidified MFA waste can be remarkably reduced by a factor of 2.4 to 2.6 by selecting NaGP or KGP, respectively, compared to BFS cement. MFA incorporation into the cementitious materials reduced the uniaxial compressive strength (UCS), though the UCS values were large enough for the final disposal in Japan.</div><div>In ultra-pure water, the Cs cumulative leached fraction (CLF) is significantly lower for geopolymers, followed by BFS cement. Although model MFA contained KCl and NaCl, geopolymers showed better performance at Cs immobilization than BFS cement. In seawater leaching, CLF of Cs exhibits no significant differences among the matrices, likely due to an increase of Cs leaching in GPs by cation exchange. Matrix choice can influence the Cs leaching behavior of solidified MFA, especially in dilute leachate environments. The Cs immobilization ability of MFA solidified with BFS cement, NaGP, and KGP met the requirement set forth by the U.S. Nuclear Regulatory Commission.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114253"},"PeriodicalIF":1.9,"publicationDate":"2025-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144518828","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

An adaptive statistical model of nuclear rods temperature for the detection of total and instantaneous blockage 用于总堵塞和瞬时堵塞检测的核棒温度自适应统计模型

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-28 DOI: 10.1016/j.nucengdes.2025.114243

Rémi Cogranne

{"title":"An adaptive statistical model of nuclear rods temperature for the detection of total and instantaneous blockage","authors":"Rémi Cogranne","doi":"10.1016/j.nucengdes.2025.114243","DOIUrl":"10.1016/j.nucengdes.2025.114243","url":null,"abstract":"<div><div>The monitoring of critical systems is of the utmost importance, especially when undetected malfunctions could lead to major accidents. This paper focuses on the temperature monitoring of fuel rod assemblies within nuclear power plants, with the goal of detecting total and instantaneous blockages as reliably and quickly as possible. First, we address the modelling of the temperature of the whole fuel rod assembly altogether. We propose a linear parametric model that is adaptive, incorporating previous temperature measurements to enhance its accuracy. This approach allows us to distinguish between regular, non-anomalous temperatures and the anomalous thermal event due to a blockage. The proposed sequential, or online, detection scheme is reliable, as the false alarm rate and detection power are analytically bounded. The model and subsequent statistical test are generic, making the methodology applicable to a wide range of nuclear cores. Numerical experiments, using real temperature measurements from the Superphénix power station, demonstrate the accuracy of the proposed model and the relevance of the detection procedure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114243"},"PeriodicalIF":1.9,"publicationDate":"2025-06-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144502319","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

Open-source microreactor design models for technoeconomic assessments 用于技术经济评估的开源微反应堆设计模型

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-28 DOI: 10.1016/j.nucengdes.2025.114210

Khaldoon Al-Dawood , Botros Hanna , Sai P. Balla , Rodrigo de Oliveira , Sam Garcia , Dan McCarthy , Chandu Bolisetti , Ben Lindley , Abdalla Abou-Jaoude

{"title":"Open-source microreactor design models for technoeconomic assessments","authors":"Khaldoon Al-Dawood , Botros Hanna , Sai P. Balla , Rodrigo de Oliveira , Sam Garcia , Dan McCarthy , Chandu Bolisetti , Ben Lindley , Abdalla Abou-Jaoude","doi":"10.1016/j.nucengdes.2025.114210","DOIUrl":"10.1016/j.nucengdes.2025.114210","url":null,"abstract":"<div><div>Technoeconomic analyses for advanced nuclear technologies are essential for identifying cost drivers which allow for optimizing the technology to achieve better economic performance. Microreactors are emerging as a promising and reliable-energy solution, offering inherent safety, low capital investment, and rapid deployment capabilities. Past studies have conducted technoeconomic analyses for microreactors, but there remains a need for open-source technoeconomic models that can enhance collaboration, transparency, and consistency when performing this type of analysis. The present article introduces an open-source technoeconomic model for microreactors based on bottom-up cost estimation. Since no real cost data for microreactors exists, this model leverages cost data and insights gleaned from the Microreactor Applications, Research, Validation and Evaluation (MARVEL) project. It estimates the first-of-a-kind cost of two microreactor technologies and can also calculate the N<sup>th</sup> of a kind (NOAK) cost via accounting for learning and mass factory production. The two technologies considered in this paper are the liquid–metal thermal reactor (LTMR) and the gas-cooled microreactor (GCMR). The goal of this study is to demonstrate bottom-up cost estimation of these microreactor technologies and provide the cost estimation models that other users can leverage for various applications such as design optimization and financial planning.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114210"},"PeriodicalIF":1.9,"publicationDate":"2025-06-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144502320","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-resolution predictions of the coolant properties for the 3D PWR core with artificial neural networks based on CTF 基于CTF的人工神经网络对3D压水堆堆芯冷却剂特性的高分辨率预测

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-28 DOI: 10.1016/j.nucengdes.2025.114261

Marianna Papadionysiou , Gregory Delipei , Maria Avramova , Hakim Ferroukhi , Kostadin Ivanov

{"title":"High-resolution predictions of the coolant properties for the 3D PWR core with artificial neural networks based on CTF","authors":"Marianna Papadionysiou , Gregory Delipei , Maria Avramova , Hakim Ferroukhi , Kostadin Ivanov","doi":"10.1016/j.nucengdes.2025.114261","DOIUrl":"10.1016/j.nucengdes.2025.114261","url":null,"abstract":"<div><div>PSI and North Carolina State University are developing a high-resolution multi-physics core solver for Pressurized Water Reactor (PWR) analysis in Cartesian geometry, using the neutron transport code nTRACER and two Machine Learning (ML) models providing thermal–hydraulic (T/H) feedback. This work focuses on the ML models, trained with CTF data to predict PWR subchannel coolant properties during normal operation. The methodology presented can be applied to different PWR core, to produce ML models capable of high-resolution T/H predictions. The ML model’s performance is evaluated on quarter-core CTF calculations achieving an average temperature difference of 1 °C from CTF and equivalent density accuracy. Their verification is extended outside configurations seen in their training, with varying mass flux, a water liner and different lattice geometry. They are also compared to a simplified one-dimensional T/H solver, showing significantly lower discrepancies, almost half, with similar computational cost, while being up to 15 times faster than CTF.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114261"},"PeriodicalIF":1.9,"publicationDate":"2025-06-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144502318","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