Pub Date : 2024-11-22DOI: 10.1016/j.nucengdes.2024.113723
M. Lovecký, J. Závorka
This study delves into the optimization of dual-purpose casks utilized for storing and transporting spent nuclear fuel from VVER-1000 reactors. Central to this investigation is the assessment of radiation dose rates surrounding the casks, facilitated by the novel OPOS-1000 calculation tool, which incorporates adjoint flux techniques for enhanced precision. This tool enables the strategic placement of hotter fuel assemblies at the core, surrounded by cooler ones, effectively minimizing radiation exposure through a carefully designed zoning strategy. Detailed analyses conducted with OPOS-1000 offer insights into the optimal configuration of fuel assemblies to reduce radiation levels, presenting a significant advancement in spent fuel cask loading efficiency. This research’s findings have the potential to streamline the loading process and influence the certification of new fuel types for existing casks, marking a pivotal step forward in spent nuclear fuel management.
{"title":"OPOS-1000: Advancing the efficiency of VVER-1000 spent nuclear fuel cask loading","authors":"M. Lovecký, J. Závorka","doi":"10.1016/j.nucengdes.2024.113723","DOIUrl":"10.1016/j.nucengdes.2024.113723","url":null,"abstract":"<div><div>This study delves into the optimization of dual-purpose casks utilized for storing and transporting spent nuclear fuel from VVER-1000 reactors. Central to this investigation is the assessment of radiation dose rates surrounding the casks, facilitated by the novel OPOS-1000 calculation tool, which incorporates adjoint flux techniques for enhanced precision. This tool enables the strategic placement of hotter fuel assemblies at the core, surrounded by cooler ones, effectively minimizing radiation exposure through a carefully designed zoning strategy. Detailed analyses conducted with OPOS-1000 offer insights into the optimal configuration of<!--> <!-->fuel assemblies to reduce radiation levels, presenting a significant advancement in spent fuel cask loading efficiency. This research’s findings have the potential to streamline the loading process and influence the certification of new fuel types for existing casks, marking a pivotal step forward in spent nuclear fuel management.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113723"},"PeriodicalIF":1.9,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701229","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}
In case of a nuclear accident with core damage in a light water reactor, the oxidation of the fuel cladding and other materials could lead to the release of combustible gases (H2 and CO) to the containment building. To mitigate the potential risk of combustion of these gases, passive autocatalytic recombiners (PARs) have been installed in numerous nuclear reactors in Europe and worldwide. PARs recombine H2 and CO with O2 producing H2O and CO2, respectively, without an open flame.
PARUPM is a code that simulates the behaviour of PARs using a physicochemical model approach. In the framework of the AMHYCO project (EU-funded Horizon 2020 project), which seeks to advance the understanding and simulation capabilities to support the combustion risk management in severe accidents, the code has been extensively enhanced and developed to simulate PAR operation with H2/CO/O2/steam mixtures. Alongside these new capabilities, the code needed a new validation process.
In this paper, the process of validation of PARUPM as a standalone code is described. The validation for steady state conditions was achieved through comparison with REKO-3 experimental data while the transient conditions were compared with results obtained with the THAI test facility. A thorough analysis of the code capabilities was performed by comparing the numerical results with experimental data for a broad series of conditions, namely: a range of different input gas temperatures and concentrations, oxygen starvation, CO poisoning, etc.
{"title":"Validation process against late phase conditions of the passive autocatalytic recombiner simulation code PARUPM as a standalone tool using experimental data from REKO-3 and THAI facilities","authors":"Araceli Dominguez-Bugarin , Ernst-Arndt Reinecke , Gonzalo Jiménez , Miguel Ángel Jiménez , Sanjeev Gupta","doi":"10.1016/j.nucengdes.2024.113722","DOIUrl":"10.1016/j.nucengdes.2024.113722","url":null,"abstract":"<div><div>In case of a nuclear accident with core damage in a light water reactor, the oxidation of the fuel cladding and other materials could lead to the release of combustible gases (H<sub>2</sub> and CO) to the containment building. To mitigate the potential risk of combustion of these gases, passive autocatalytic recombiners (PARs) have been installed in numerous nuclear reactors in Europe and worldwide. PARs recombine H<sub>2</sub> and CO with O<sub>2</sub> producing H<sub>2</sub>O and CO<sub>2</sub>, respectively, without an open flame.</div><div>PARUPM is a code that simulates the behaviour of PARs using a physicochemical model approach. In the framework of the AMHYCO project (EU-funded Horizon 2020 project), which seeks to advance the understanding and simulation capabilities to support the combustion risk management in severe accidents, the code has been extensively enhanced and developed to simulate PAR operation with H<sub>2</sub>/CO/O<sub>2</sub>/steam mixtures. Alongside these new capabilities, the code needed a new validation process.</div><div>In this paper, the process of validation of PARUPM as a standalone code is described. The validation for steady state conditions was achieved through comparison with REKO-3 experimental data while the transient conditions were compared with results obtained with the THAI test facility. A thorough analysis of the code capabilities was performed by comparing the numerical results with experimental data for a broad series of conditions, namely: a range of different input gas temperatures and concentrations, oxygen starvation, CO poisoning, etc.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113722"},"PeriodicalIF":1.9,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701231","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}
Pub Date : 2024-11-20DOI: 10.1016/j.nucengdes.2024.113712
R.T. Sweet , C.P. Massey , J.A. Hirschhorn , S.B. Bell , K.A. Kane
Cladding burst experiments for FeCrAl cladding were performed in the Severe Accident Test Station facility at Oak Ridge National Laboratory. These experiments were simulated using the BISON fuel performance code to better understand the cladding plastic behavior and failure under simulated loss-of-coolant accident conditions. 3D cladding surface boundary conditions were generated using composite axial and azimuthal profiles from experiment thermocouple data. To improve the simulation analysis capabilities in BISON for cladding burst behavior, new thermal creep, plasticity, and failure stress models specific to C26M, a wrought FeCrAl alloy, were developed and implemented.
Initial cladding burst results indicated a general underprediction in the failure temperature of the six cladding burst simulations versus the observed failure temperatures. Close investigation of the experiment timing versus the underlying tensile test data revealed that, compared with the tensile specimens, the cladding tubes did not experience the same long holding time at high temperatures. New tensile tests were performed at high temperatures using a temperature ramp similar to the simulated loss-of-coolant accident experiments. These new tensile curves showed an approximately 80% increase in the ultimate tensile strength of the C26M alloy, indicating that a holding time of 10 min at 700 °C and 800 °C allows annealing to change the material microstructure.
Using the updated tensile properties, the burst temperatures and stresses from the simulations showed remarkable agreement with the experimental results. This study was then extended by varying the initial pressure to highlight the burst temperature difference between standard Zircaloy-4 and C26M cladding under equivalent conditions. The results show that C26M has a burst temperature that is approximately 70–130 K greater than that of Zircaloy-4.
These modeling predictions can be further improved by collecting high-temperature tensile data for C26M beyond the temperature ranges used in this work.
{"title":"Wrought FeCrAl alloy (C26M) cladding behavior and burst under simulated loss-of-coolant accident conditions","authors":"R.T. Sweet , C.P. Massey , J.A. Hirschhorn , S.B. Bell , K.A. Kane","doi":"10.1016/j.nucengdes.2024.113712","DOIUrl":"10.1016/j.nucengdes.2024.113712","url":null,"abstract":"<div><div>Cladding burst experiments for FeCrAl cladding were performed in the Severe Accident Test Station facility at Oak Ridge National Laboratory. These experiments were simulated using the BISON fuel performance code to better understand the cladding plastic behavior and failure under simulated loss-of-coolant accident conditions. 3D cladding surface boundary conditions were generated using composite axial and azimuthal profiles from experiment thermocouple data. To improve the simulation analysis capabilities in BISON for cladding burst behavior, new thermal creep, plasticity, and failure stress models specific to C26M, a wrought FeCrAl alloy, were developed and implemented.</div><div>Initial cladding burst results indicated a general underprediction in the failure temperature of the six cladding burst simulations versus the observed failure temperatures. Close investigation of the experiment timing versus the underlying tensile test data revealed that, compared with the tensile specimens, the cladding tubes did not experience the same long holding time at high temperatures. New tensile tests were performed at high temperatures using a temperature ramp similar to the simulated loss-of-coolant accident experiments. These new tensile curves showed an approximately 80% increase in the ultimate tensile strength of the C26M alloy, indicating that a holding time of 10 min at 700 °C and 800 °C allows annealing to change the material microstructure.</div><div>Using the updated tensile properties, the burst temperatures and stresses from the simulations showed remarkable agreement with the experimental results. This study was then extended by varying the initial pressure to highlight the burst temperature difference between standard Zircaloy-4 and C26M cladding under equivalent conditions. The results show that C26M has a burst temperature that is approximately 70–130 K greater than that of Zircaloy-4.</div><div>These modeling predictions can be further improved by collecting high-temperature tensile data for C26M beyond the temperature ranges used in this work.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113712"},"PeriodicalIF":1.9,"publicationDate":"2024-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701264","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}
Pub Date : 2024-11-20DOI: 10.1016/j.nucengdes.2024.113703
N. Seiler , M. Johnson , L. Vyskocil , Y. Vorobyov , W. Villanueva , M. Abu Bakar , O. Zhabin , M. Kratochvil , B. Bian , A. Drouillet
Convection within shallow pools of liquid metals heated from below is of significant interest for the In-Vessel Retention (IVR) strategy for Pressurised Water Reactors (PWR) as focusing of the lateral heat flux at the reactor wall presents a risk to the thermomechanical integrity of the reactor vessel. Under an IAEA Coordinated Research Project on corium melt retention, various international research institutions have performed CFD simulations to predict the thermal–hydraulic behaviour of a prototypic light metal layer of low Prandtl number () and high external Rayleigh number () dissipating heat from the free surface and at the lateral reactor wall. Various computational approaches including LES-WALE, LES-Smagorinsky and spectral-DNS were validated under the conditions of two BALI-Metal experiments in water (), revealing promising agreement in the predicted repartition of the heat flux at the vertical and lateral boundaries. Simulations in a prototypic light metal layer indicated 30–34 % of heat dissipation due to thermal radiation at the free surface. Average thermal losses at the lateral wall corresponded to a focusing effect of 3.3–3.7 times the imposed heat flux. A spike in lateral heat flux close to the free surface equated to a local focusing effect 6-times the imposed heat flux from below. The fluid dynamics, driven largely by thermal losses at the reactor wall, were characterised by downwards acceleration adjacent to the lateral wall and ejection of a cold jet parallel to the lower boundary, forming a large convection cell comparable in size to the radius of the reactor.
{"title":"Natural convection in a shallow pool heated from below and implications for the thermal focusing effect at the lateral wall","authors":"N. Seiler , M. Johnson , L. Vyskocil , Y. Vorobyov , W. Villanueva , M. Abu Bakar , O. Zhabin , M. Kratochvil , B. Bian , A. Drouillet","doi":"10.1016/j.nucengdes.2024.113703","DOIUrl":"10.1016/j.nucengdes.2024.113703","url":null,"abstract":"<div><div>Convection within shallow pools of liquid metals heated from below is of significant interest for the In-Vessel Retention (IVR) strategy for Pressurised Water Reactors (PWR) as focusing of the lateral heat flux at the reactor wall presents a risk to the thermomechanical integrity of the reactor vessel. Under an IAEA Coordinated Research Project on corium melt retention, various international research institutions have performed CFD simulations to predict the thermal–hydraulic behaviour of a prototypic light metal layer of low Prandtl number (<span><math><mrow><mi>Pr</mi><mo>=</mo><mn>0.02</mn></mrow></math></span>) and high external Rayleigh number (<span><math><mrow><mi>R</mi><msub><mi>a</mi><mi>Φ</mi></msub><mo>∼</mo><msup><mrow><mn>10</mn></mrow><mn>12</mn></msup></mrow></math></span>) dissipating heat from the free surface and at the lateral reactor wall. Various computational approaches including LES-WALE, LES-Smagorinsky and spectral-DNS were validated under the conditions of two BALI-Metal experiments in water (<span><math><mrow><mi>Pr</mi><mo>=</mo><mn>6.9</mn></mrow></math></span>), revealing promising agreement in the predicted repartition of the heat flux at the vertical and lateral boundaries. Simulations in a prototypic light metal layer indicated 30–34 % of heat dissipation due to thermal radiation at the free surface. Average thermal losses at the lateral wall corresponded to a focusing effect of 3.3–3.7 times the imposed heat flux. A spike in lateral heat flux close to the free surface equated to a local focusing effect 6-times the imposed heat flux from below. The fluid dynamics, driven largely by thermal losses at the reactor wall, were characterised by downwards acceleration adjacent to the lateral wall and ejection of a cold jet parallel to the lower boundary, forming a large convection cell comparable in size to the radius of the reactor.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113703"},"PeriodicalIF":1.9,"publicationDate":"2024-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701228","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}
Pub Date : 2024-11-19DOI: 10.1016/j.nucengdes.2024.113715
Chao Yuan , Tianbao Lan
Nuclear graphite is an ideal material for neutron moderators and reflectors of nuclear power systems that bear a severe environment of high temperature (up to 1000 ℃) and accumulated neutron irradiation (up to 1027n/m2). In service, the mechanical properties of the nuclear graphite considerably evolve with the unsteady coupled thermal-irradiation field, bringing out undesired internal stresses and deformations that potentially imperil the structural integrity and reliability. Although there is a large temperature gradient and nonlinear irradiation distribution in real working conditions, the majority of the open literature does not take these effects into consideration during the stress analysis. Herein, in order to provide a safe assessment of the structural integrity, we take the cylindrical IG-110 nuclear graphite reflector as a representative to numerically investigate the effects of temperature and irradiation gradient on the temporal and spatial variations of the stress field. Numerical analysis indicates that regardless of the magnitude of the temperature gradient and irradiation gradient, the maximum tensile stress of the whole structure is always achieved after fixed periods of operation and is located at the inner surface of the cylinder. However, a greater maximum tensile stress can be induced under an inhomogeneous temperature field of larger gradients, or a nonlinear irradiation field of smaller gradient factors. Compared with conventional analyses that ignore the effect of the thermal-irradiation gradient, our analysis renders a safe and conservative design for nuclear graphite structures.
{"title":"Effects of temperature gradient and nonlinear neutron irradiation on the stress in nuclear graphite reflector","authors":"Chao Yuan , Tianbao Lan","doi":"10.1016/j.nucengdes.2024.113715","DOIUrl":"10.1016/j.nucengdes.2024.113715","url":null,"abstract":"<div><div>Nuclear graphite is an ideal material for neutron moderators and reflectors of nuclear power systems that bear a severe environment of high temperature (up to 1000 ℃) and accumulated neutron irradiation (up to 10<sup>27</sup>n/m<sup>2</sup>). In service, the mechanical properties of the nuclear graphite considerably evolve with the unsteady coupled thermal-irradiation field, bringing out undesired internal stresses and deformations that potentially imperil the structural integrity and reliability. Although there is a large temperature gradient and nonlinear irradiation distribution in real working conditions, the majority of the open literature does not take these effects into consideration during the stress analysis. Herein, in order to provide a safe assessment of the structural integrity, we take the cylindrical IG-110 nuclear graphite reflector as a representative to numerically investigate the effects of temperature and irradiation gradient on the temporal and spatial variations of the stress field. Numerical analysis indicates that regardless of the magnitude of the temperature gradient and irradiation gradient, the maximum tensile stress of the whole structure is always achieved after fixed periods of operation and is located at the inner surface of the cylinder. However, a greater maximum tensile stress can be induced under an inhomogeneous temperature field of larger gradients, or a nonlinear irradiation field of smaller gradient factors. Compared with conventional analyses that ignore the effect of the thermal-irradiation gradient, our analysis renders a safe and conservative design for nuclear graphite structures.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113715"},"PeriodicalIF":1.9,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701263","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}
Pub Date : 2024-11-19DOI: 10.1016/j.nucengdes.2024.113686
Lei Liu , Xiuhua Chen , Junhao Gao , Chenyu Zhou , Lijun Liu
The vortex structures may cause unstable pressure pulsations and vibrations in Reactor Coolant Pumps (RCPs). During the CAP1400 RCP tests, flow-induced vibrations were observed at both the inherent frequency (3.66 Hz) and the blade passing frequency (fBPF). The amplitude of the vibrations at the inherent frequency exceeded the allowable limit, resulting in necessary shutdown inspections. This study, based on field test data, thoroughly analyzed the characteristics of low-frequency (below 10 Hz) pressure pulsations in the complex experimental loop and identified the main occurrence area as the outlet of the RCP and the inlet of the valve. We also briefly analyzed the amplitude variations of the fBPF and 2fBPF in the experimental loop, observing similar periodic variations at specific frequencies under different rotational speeds. Additionally, as the rotational speed increased, the amplitudes of both the low-frequency and fBPF pressure pulsations significantly increased. Numerical simulations revealed the flow field within the experimental loop. The interaction between the flow from the RCP diffuser outlet and the volute, as well as the sudden change in the flow channel area within the valve, are the main mechanisms forming vortices. These findings provide important test and theoretical foundations for further studies on flow-induced vibration problems in mixed-flow RCPs and experimental loops.
{"title":"Pressure pulsation analysis of a large reactor coolant pump experimental loop based on field test and numerical simulation","authors":"Lei Liu , Xiuhua Chen , Junhao Gao , Chenyu Zhou , Lijun Liu","doi":"10.1016/j.nucengdes.2024.113686","DOIUrl":"10.1016/j.nucengdes.2024.113686","url":null,"abstract":"<div><div>The vortex structures may cause unstable pressure pulsations and vibrations in Reactor Coolant Pumps (RCPs). During the CAP1400 RCP tests, flow-induced vibrations were observed at both the inherent frequency (3.66 Hz) and the blade passing frequency (<em>f</em><sub>BPF</sub>). The amplitude of the vibrations at the inherent frequency exceeded the allowable limit, resulting in necessary shutdown inspections. This study, based on field test data, thoroughly analyzed the characteristics of low-frequency (below 10 Hz) pressure pulsations in the complex experimental loop and identified the main occurrence area as the outlet of the RCP and the inlet of the valve. We also briefly analyzed the amplitude variations of the <em>f</em><sub>BPF</sub> and 2<em>f</em><sub>BPF</sub> in the experimental loop, observing similar periodic variations at specific frequencies under different rotational speeds. Additionally, as the rotational speed increased, the amplitudes of both the low-frequency and <em>f</em><sub>BPF</sub> pressure pulsations significantly increased. Numerical simulations revealed the flow field within the experimental loop. The interaction between the flow from the RCP diffuser outlet and the volute, as well as the sudden change in the flow channel area within the valve, are the main mechanisms forming vortices. These findings provide important test and theoretical foundations for further studies on flow-induced vibration problems in mixed-flow RCPs and experimental loops.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113686"},"PeriodicalIF":1.9,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701261","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}
Pub Date : 2024-11-19DOI: 10.1016/j.nucengdes.2024.113721
Xingfang Wang, Youqi Zheng, Xiayu Wang, Xiaoqi Li
Ex-core detectors have the potential to monitor the power distribution in small space reactors. However, there are still considerable challenges remain in their practical implementation. To address this gap, this paper proposes a novel method to monitor power distribution utilizing ex-core detectors and deep neural networks. A small space reactor model simplified from TOPAZ-II was constructed based on the assumption that 12 ex-core detectors could be applied. New neural network models were established to consider the differences of pin power at different positions in the core by independently modeling the inner and outer fuel pins. This method was extensively validated across a wide range of operational conditions. The deep neural network method also exhibits reduced sensitivity to noise. By training on datasets containing noisy signals, the neural network method can handle signals containing ± 1 % noise while the accuracy of power distribution predictions is maintained. In addition, the deep neural network method is capable of monitoring asymmetric power distribution. By learning the characteristics of signals from asymmetric detectors, this method can accurately predict core power distribution even under abnormal operational conditions.
{"title":"Using ex-core detectors and deep neural networks for monitoring power distribution in small space reactors","authors":"Xingfang Wang, Youqi Zheng, Xiayu Wang, Xiaoqi Li","doi":"10.1016/j.nucengdes.2024.113721","DOIUrl":"10.1016/j.nucengdes.2024.113721","url":null,"abstract":"<div><div>Ex-core detectors have the potential to monitor the power distribution in small space reactors. However, there are still considerable challenges remain in their practical implementation. To address this gap, this paper proposes a novel method to monitor power distribution utilizing ex-core detectors and deep neural networks. A small space reactor model simplified from TOPAZ-II was constructed based on the assumption that 12 ex-core detectors could be applied. New neural network models were established to consider the differences of pin power at different positions in the core by independently modeling the inner and outer fuel pins. This method was extensively validated across a wide range of operational conditions. The deep neural network method also exhibits reduced sensitivity to noise. By training on datasets containing noisy signals, the neural network method can handle signals containing ± 1 % noise while the accuracy of power distribution predictions is maintained. In addition, the deep neural network method is capable of monitoring asymmetric power distribution. By learning the characteristics of signals from asymmetric detectors, this method can accurately predict core power distribution even under abnormal operational conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113721"},"PeriodicalIF":1.9,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701262","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}
Pub Date : 2024-11-19DOI: 10.1016/j.nucengdes.2024.113719
Damian Christopher Selvam , Yuvarajan Devarajan , T. Raja
This study examines the increasing potential of artificial intelligence (AI) to transform nuclear waste management by enhancing procedures related to waste classification, treatment, storage, and disposal. The ability of AI to examine extensive datasets via machine learning and data analytics improves the accuracy and efficiency of trash classification. Additionally, AI-driven optimization methods enhance treatment procedures, reduce risks, and guarantee adherence to stringent regulatory standards, resulting in safer management of radioactive materials. These developments in AI enhance operational efficiency and refine decision-making frameworks, facilitating more accurate risk assessments. Integrating AI into nuclear waste management enables stakeholders to negotiate intricate regulatory frameworks more efficiently while minimizing environmental consequences and safeguarding public health.
This report identifies essential domains for forthcoming research and development in AI-augmented nuclear waste management. Essential directives encompass enhancing AI algorithms for real-time surveillance and predictive analytics, facilitating the early identification of possible problems, and enabling more proactive management. Moreover, developing technologies like robotic systems and autonomous platforms possess the capability to automate numerous waste management jobs, hence diminishing human risk exposure. The continuous developments illustrate AI’s revolutionary capacity to tackle critical issues in nuclear waste management, guaranteeing the safe, responsible, and sustainable management of radioactive materials for future generations.
{"title":"Exploring the potential of artificial intelligence in nuclear waste management: Applications, challenges, and future directions","authors":"Damian Christopher Selvam , Yuvarajan Devarajan , T. Raja","doi":"10.1016/j.nucengdes.2024.113719","DOIUrl":"10.1016/j.nucengdes.2024.113719","url":null,"abstract":"<div><div>This study examines the increasing potential of artificial intelligence (AI) to transform nuclear waste management by enhancing procedures related to waste classification, treatment, storage, and disposal. The ability of AI to examine extensive datasets via machine learning and data analytics improves the accuracy and efficiency of trash classification. Additionally, AI-driven optimization methods enhance treatment procedures, reduce risks, and guarantee adherence to stringent regulatory standards, resulting in safer management of radioactive materials. These developments in AI enhance operational efficiency and refine decision-making frameworks, facilitating more accurate risk assessments. Integrating AI into nuclear waste management enables stakeholders to negotiate intricate regulatory frameworks more efficiently while minimizing environmental consequences and safeguarding public health.</div><div>This report identifies essential domains for forthcoming research and development in AI-augmented nuclear waste management. Essential directives encompass enhancing AI algorithms for real-time surveillance and predictive analytics, facilitating the early identification of possible problems, and enabling more proactive management. Moreover, developing technologies like robotic systems and autonomous platforms possess the capability to automate numerous waste management jobs, hence diminishing human risk exposure. The continuous developments illustrate AI’s revolutionary capacity to tackle critical issues in nuclear waste management, guaranteeing the safe, responsible, and sustainable management of radioactive materials for future generations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113719"},"PeriodicalIF":1.9,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701179","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}
Pub Date : 2024-11-19DOI: 10.1016/j.nucengdes.2024.113714
Tianbao Lan , Xingming Peng , FengSheng , Wei Tan
To evaluate the structural safety of the graphite core in a gas-cooled micro-reactor and to assess its structural response under seismic loads, a study was conducted. By comparing the acceleration and velocity curves obtained from small-sized graphite block collision experiments and collision simulations, it was determined that the simulation results accurately represent the real collision behavior of graphite blocks. The collision stiffness and damping parameters were derived from these curves. Subsequently, simulations of graphite components in the core were performed to establish the stiffness and damping parameters of the graphite blocks, which were then incorporated into the core analysis calculations. To validate the accuracy of the core numerical model and simplify the vibration form, the core model was divided into in-plane and axial models. A full-core model calculation was then carried out to determine the forces between graphite components. The final results confirm that the graphite core adheres to the ASME design specifications under seismic loads.
{"title":"Seismic modeling and simulation of the graphite core in gas-cooled micro-reactor","authors":"Tianbao Lan , Xingming Peng , FengSheng , Wei Tan","doi":"10.1016/j.nucengdes.2024.113714","DOIUrl":"10.1016/j.nucengdes.2024.113714","url":null,"abstract":"<div><div>To evaluate the structural safety of the graphite core in a gas-cooled micro-reactor and to assess its structural response under seismic loads, a study was conducted. By comparing the acceleration and velocity curves obtained from small-sized graphite block collision experiments and collision simulations, it was determined that the simulation results accurately represent the real collision behavior of graphite blocks. The collision stiffness and damping parameters were derived from these curves. Subsequently, simulations of graphite components in the core were performed to establish the stiffness and damping parameters of the graphite blocks, which were then incorporated into the core analysis calculations. To validate the accuracy of the core numerical model and simplify the vibration form, the core model was divided into in-plane and axial models. A full-core model calculation was then carried out to determine the forces between graphite components. The final results confirm that the graphite core adheres to the ASME design specifications under seismic loads.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113714"},"PeriodicalIF":1.9,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142701180","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}
Pub Date : 2024-11-18DOI: 10.1016/j.nucengdes.2024.113717
M. Sharifi, M. Aghaie
In this research, the COMSOL is used as a multi-physics software to simulate the thermo-mechanical performance of a UO2-SiC fuel rod in hypothetical accidents of WWERs. First, by using the thermal and mechanical analysis, the temperature, strain and stress in different parts of the UO2-SiC fuel rod in normal operation are calculated. Next the performance of fuel rod in hypothetical reactivity insertions and hypothetical semi loss of coolant is evaluated. The results of the thermal and mechanical analysis of UO2-SiC fuel with different percentages of SiC and the substitution of Si instead of zirconium clad are also analyzed. Coupling point kinetic equations with COMSOL events related to the increase in reactor power, such as the reactivity insertion accident (RIA), are simulated. For more detail study, hydrogen diffusion in the clad, oxygen diffusion in the fuel and non-stoichiometric fuels are considered. Finally, a sensitivity analysis is carried out to see how the effective quantities affect the mechanical and thermal evaluations and COMSOL is introduced as a multi physics software could present accident evaluations.
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