Nuclear Engineering and Design最新文献

{"title":"Modelica and Arduino-based hardware-in-the-loop simulation of a nuclear-powered engineering ship","authors":"Ao Zhang ,&nbsp;Xun He ,&nbsp;Antonio Cammi ,&nbsp;Xiang Wang","doi":"10.1016/j.nucengdes.2024.113650","DOIUrl":"10.1016/j.nucengdes.2024.113650","url":null,"abstract":"<div><div>Hardware-in-the-loop simulation (HILS) can enhance the authenticity and reliability of offline simulation at the design stage, and reduce the risk of equipment performance commissioning for complex systems. This paper focuses on the HILS of the control mechanism of a nuclear-powered engineering ship including a two-loop nuclear power system, an electric power system, and a mechanical system. In the software part, the mathematical model is implemented in Modelica language with OpenModelica. The nuclear power system is demonstrated in steady-state referring to the Japanese nuclear power merchant ship “NS Mutsu” and is debugged under transient conditions with the help of an operation control module. The electric power system and mechanical system were developed to have certain functions for marine engineering. In the hardware part, the system is built based on an Arduino microcontroller, the Modelica open-source library, and a Bluetooth-based communication protocol between the computer and the microcontroller. The study proved that HILS is capable of simulating the multi-physical joint operation on the software level, establishing the real-time action response and data feedback between the software and hardware parts, and completing steady-state as well as various transient simulations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535521","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":"Applications of deep reinforcement learning in nuclear energy: A review","authors":"Yongchao Liu ,&nbsp;Bo Wang ,&nbsp;Sichao Tan ,&nbsp;Tong Li ,&nbsp;Wei Lv ,&nbsp;Zhenfeng Niu ,&nbsp;Jiangkuan Li ,&nbsp;Puzhen Gao ,&nbsp;Ruifeng Tian","doi":"10.1016/j.nucengdes.2024.113655","DOIUrl":"10.1016/j.nucengdes.2024.113655","url":null,"abstract":"<div><div>In recent years, Deep reinforcement learning (DRL), as an important branch of artificial intelligence (AI), has been widely used in physics and engineering domains. It combines the perceptual advantages of deep learning (DL) and the decision-making advantages of reinforcement learning (RL), and is very suitable for solving the “perception-decision” problem with high-dimensional and nonlinear characteristics. In this paper, firstly, the algorithm principle, mainstream framework, characteristics and advantages of DRL are summarized. Secondly, the application research status of DRL in other energy fields is reviewed, which provides reference for the possible impact and future research direction in the field of nuclear energy. Thirdly, the main research directions of DRL in the field of nuclear energy are summarized and commented, and the application architecture and advantages of DRL are illustrated through specific application cases. Finally, the advantages, limitations and future development direction of DRL in the field of nuclear energy are discussed. The goal of this review is to provide an understanding of DRL capabilities along with state-of-the-art applications in nuclear energy to researchers wishing to address new problems with these methods.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535545","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":"Characterization of long-term evolution of leakage rates of O-ring seals in nuclear power plant under high-temperature and high-pressure conditions","authors":"Ming Li ,&nbsp;Lei He ,&nbsp;Dengke Zheng ,&nbsp;Xing Fang ,&nbsp;Xiaoming Huang ,&nbsp;Guoliang Xu","doi":"10.1016/j.nucengdes.2024.113666","DOIUrl":"10.1016/j.nucengdes.2024.113666","url":null,"abstract":"<div><div>Under accident conditions, the sustained high temperature and high pressure (HTHP) environment within the containment structure poses a threat to the penetration seals. This study integrates finite element analysis of the mechanical properties of O-ring, both hyperelastic and viscoelastic characteristics, and an interfacial leakage model to predict the variation in O-ring leakage rates over time in HTHP environments. A series of leakage test experiments are conducted to validate the predictive model, indicating good agreement between experimental and predicted values. The effects of HTHP on non-aged O-rings (short-term service) are analyzed through mechanical simulations and leakage rate calculations. The results reveal that high temperatures positively and reversibly affect the O-ring seals, with hazardous conditions mainly resulting from over-pressurization. However, during the long-term service of aged O-rings, thermal aging caused by high temperatures significantly influences leakage rates. The thermal aging coupled with high pressure can cause material damage (such as rubber being squeezed out) and functional failures (excessive leakage rates). The long-term leakage rates of O-rings at high temperatures in further investigation exhibits a time–temperature equivalence. The master curve is plotted to derive an equation that describes the relationship between leakage rates, temperature, and time under specific pressure conditions. The equation indicates that the dimensionless leakage rate serves as an indicator of seal degradation and enables the quantitative evaluation of the long-term service life of the O-rings using the maximum allowable leakage rate. These findings are applicable within the range of accidental operating conditions for containment structures, including temperatures up to 160 °C and pressures up to 0.75 MPa.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535520","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":"Study and analysis of the effect of dynamic load on the application of LBB technology to the primary pipe of reactor coolant system","authors":"Feng He ,&nbsp;Furui Xiong,&nbsp;Xiaoming Bai,&nbsp;Kang Yang,&nbsp;Xifeng Lu,&nbsp;Bingjin Li,&nbsp;Xinjun Wang","doi":"10.1016/j.nucengdes.2024.113646","DOIUrl":"10.1016/j.nucengdes.2024.113646","url":null,"abstract":"<div><div>Reactors may be subjected to dynamic loads, and in order to ensure that Leak-Before-Break (LBB) technology is safely and reliably applied to the primary piping of reactor coolant system, dynamic load impact analysis of LBB technology applied to the primary piping was performed. The dynamic mechanical properties test was carried out on the primary piping, and the crack leakage rate test under dynamic load and the impact load loading test under different loading speeds were carried out on the primary piping test piece with circumferential penetrating crack. Based on experimental data, the crack leakage rate analysis method and crack stability analysis method in quasi-static LBB technology were used to analyze the impact of applying LBB technology to the primary piping under dynamic load. The analysis results indicate that, when LBB technology is applied to the primary piping of reactor coolant system, the analysis results under dynamic load are more conservative than those under quasi-static load, and it is safer and more reliable under dynamic loads in practical engineering applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535517","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":"Generic material irradiation database for delayed heating calculations","authors":"Roberto E. Fairhurst-Agosta,&nbsp;Tomasz Kozlowski","doi":"10.1016/j.nucengdes.2024.113634","DOIUrl":"10.1016/j.nucengdes.2024.113634","url":null,"abstract":"<div><div>Safety analyses in research reactors require the estimation of heat deposited in experiments and reactor structures after shutdown. The accurate assessment of the deposited energy across a reactor geometry better determines the heat removal requirements and ensures effective cooling. However, the development of experiment safety analyses on a case-by-case basis often proves to be effort and time-consuming. This article introduces a method based on the creation of a generic material irradiation database to expedite the process. The irradiation database is created by calculating the delayed heating in experiments of individual chemical elements. Then, the created database enables the quick calculation of the delayed heating of experiments of arbitrary material composition. This article showcases two applications to demonstrate the delayed heating calculation workflow and verify the generic material irradiation database method. These applications include a simple demonstration exercise and an Advanced Test Reactor experiment. The results display an overall good agreement between the generic material irradiation database method and reference values for a wide variety of experiments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535519","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":"Investigation on nonlinear dynamics and fatigue damage of a simply supported tube with clearance restriction in tube bundles subjected to two-phase flow","authors":"Shihao Yang ,&nbsp;Jiang Lai","doi":"10.1016/j.nucengdes.2024.113644","DOIUrl":"10.1016/j.nucengdes.2024.113644","url":null,"abstract":"<div><div>In this study, a mathematical model was developed to analyze the vibration of a simply supported tube with clearance restriction in an otherwise rigid rotated triangular tube array subjected to two-phase flow. The study evaluated the fatigue damage and collision behavior of the tube bundles system. The numerical results showed that the reduced velocity significantly affects the dominant frequency, collision frequency, and stable equilibrium position of the system. This suggests that the nonlinear dynamic characteristics of the simply supported tube with clearance restriction play a crucial role in determining fatigue damage.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535544","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":"COAL reflooding experiments during a loss of coolant Accident: Effect of the water flow rate, the pressure and the rod power with ballooned rods","authors":"G. Repetto ,&nbsp;Q. Grando ,&nbsp;S. Eymery ,&nbsp;R. Van Lochem","doi":"10.1016/j.nucengdes.2024.113641","DOIUrl":"10.1016/j.nucengdes.2024.113641","url":null,"abstract":"<div><div>During a loss of coolant accident (LOCA) in a pressurized water reactor, the drying of the fuel assemblies leads to an increase in the fuel temperature and a deformation of the fuel rod cladding.</div><div>The COAL experiments focused on the coolability issue of a partially deformed fuel assembly during water injection with the safety systems using a 7x7 bundle of electrically heated rods. The relocation of the fragmented fuel in the balloons is taken into account by a local increase in power by a factor of 1.5, and the effect of the flow area restriction is provided with various flow blockage (intact geometry up to moderate and long ballooning (100 and 300 mm) with different blockage ratios (80 and 90 %)).</div><div>These experiments, in the frame of the PERFROI project, were launched by the “Institut de Radioprotection et de Sureté Nucléaire” (IRSN)<del>.</del></div><div>This paper presents the thermal hydraulics parameters and the main results of some experiments carried out in a facility of the STERN Laboratories. We studied the effect of the inlet water flow rate which is the consequence of the amount of water entering the reactor core after the break of the primary circuit, the effect of the pressure and the effect of the rod power as a function of the moment of availability of the safety pumps after the reactor scram. We provide experiments data on the coolability limits for different rod powers, which is given by the minimum of water flow to consider that the reflooding may be not impaired (PCT below the LOCA criterium of 1204 °C). The needed flow is ranging from 7 7 kg/s/m² (with intact rods geometry) at low power up 35 kg/s/m² (with at the high power that remaining in the core 1 min after the reactor scram) with a strong effect of the presence a partially local area due to rod ballooning during the large break LOCA accident. We outlined also the effect of the system pressure with a strong effect on the reflooding process above 10 bar up to 30 (for medium break LOCA).</div><div>These results are used to improve and validate the heat exchange models of thermal hydraulics codes dealing with the complex reflooding processes in such a configuration.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535518","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 the ENIGMA fuel performance code for LWR applications with chromium-coated cladding","authors":"Glyn Rossiter ,&nbsp;Kerr Fitzgerald ,&nbsp;Aiden Peakman","doi":"10.1016/j.nucengdes.2024.113656","DOIUrl":"10.1016/j.nucengdes.2024.113656","url":null,"abstract":"<div><div>Zirconium-alloy cladding with a chromium coating is the most advanced of the near-term concepts amongst Accident Tolerant Fuel (ATF) materials for Light Water Reactor (LWR) applications. The ENIGMA fuel performance code has been updated to model the thermo-mechanical behaviour of such cladding in both normal and off-normal operating conditions. The focus was on accurately simulating the behaviour in Loss Of Coolant Accident (LOCA) conditions to evaluate the increase in coping time during design-basis accidents. New low-temperature and high-temperature models were incorporated for cladding oxidation and hydriding and cladding creep which take into account the impact of the chromium coating on the overall cladding behaviour. Furthermore, the consumption of the chromium coating due to high-temperature diffusion of chromium into the cladding base alloy’s β-Zr phase is simulated. The new models have been validated using measurements on chromium-coated cladding from irradiated rods, high-temperature annealing experiments and semi-integral LOCA tests. The validation showed good agreement between ENIGMA’s predictions and the experimental data; thereby demonstrating the applicability of the new models for simulating the performance of LWR fuel rods with chromium-coated cladding in both normal operation and accident conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535515","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":"Preliminary analysis of the in-orbit operation characteristic of the inherent safety space reactor power system","authors":"Li Huaqi,&nbsp;Tian Xiaoyan,&nbsp;Wei Mingyan,&nbsp;Zhu Lei,&nbsp;Shi Leitai,&nbsp;Chen Sen,&nbsp;Luo Xiaofei,&nbsp;Li Da,&nbsp;Chen Lixin,&nbsp;Jiang Xinbiao","doi":"10.1016/j.nucengdes.2024.113652","DOIUrl":"10.1016/j.nucengdes.2024.113652","url":null,"abstract":"<div><div>The inherent safety space reactor power system with a coupled thermoelectric conversion in a liquid metal lithium-cooled reactor represents a highly reliable space power. Compared with the ground reactor, the SNRPS has its own characteristics in safety considerations, mainly manifested in the SNRPS before launch, during launch and during the ascent into orbit will be affected by the launch vehicle. Which can be analyzed by the common methodology of probabilistic risk management. To investigate the response characteristics during in-orbit operation accidents, a transient analysis model of the liquid metal-cooled space reactor power system is established. The system response characteristics of the inherent safety space reactor power system conceptual designs are preliminarily analyzed under four potential typical in-orbit operating conditions, including (1) rated operating condition, (2) control drum misoperation events, (3) partial loss of coolant flow accident, and (4) partial failure of the radiator area accident. The results show that the power system has inherent safety in-orbit operation characteristics due to the system design operating parameters, which the coolant temperature below 1200 K at the rated operating condition. Even under typical operating accidents, the system coolant remains highly supercooled (more than 200 K), preventing boiling from occurring. The maximum temperature of the core fuel pin and cladding materials remains lower than their safety limits, ensuring that no core melting phenomenon occurs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535516","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 review of irradiation-induced hardening in FeCrAl alloy systems for accident-tolerant fuel cladding","authors":"Abylay Tangirbergen ,&nbsp;Nurlan Amangeldi ,&nbsp;Shripad T. Revankar ,&nbsp;Gani Yergaliuly","doi":"10.1016/j.nucengdes.2024.113659","DOIUrl":"10.1016/j.nucengdes.2024.113659","url":null,"abstract":"<div><div>Despite nuclear energy being a clean, sustainable source, its safety is a major concern, especially after the Chernobyl and Fukushima accidents. Designing accident-tolerant fuel (ATF) clad materials is a key solution. This review examines the development and behavior of FeCrAl alloys, a promising ATF cladding candidate, under irradiation. FeCrAl alloys show excellent resistance to high-temperature corrosion and oxidation, but irradiation can significantly alter their mechanical properties. This paper consolidates experimental and theoretical studies on irradiation hardening in FeCrAl alloys, highlighting dislocation loops and Cr-rich α’ precipitates as primary hardening contributors. It discusses compositional adjustments, such as adding oxide dispersion strengthening (ODS) materials, and evaluates advanced techniques to mitigate irradiation-induced damage and improve alloy performance. Theoretical frameworks of irradiation hardening and computer simulation methods are overviewed. This review provides a comprehensive understanding of irradiation hardening mechanisms in FeCrAl alloys and suggests future research directions for enhancing nuclear reactor safety and efficiency.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535543","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}