Pub Date : 2025-02-01DOI: 10.1016/j.pnucene.2025.105605
Seung Geun Kim , Seunghyoung Ryu , Kyungho Jin , Hyeonmin Kim
The rapid advancement of artificial intelligence (AI) technology based on deep neural networks (DNNs) has spurred active development of DNN-based models in the nuclear domain. Due to the black-box nature of these models and the issue of low explainability, their practical application in safety-critical domains is hindered. To address this, numerous explainable AI (XAI) methods have been proposed. However, the selection of an appropriate XAI method is crucial as its performance significantly depends on various factors; nonetheless, comparative studies of XAI methods are limited within the nuclear domain. This study employs perturbation analysis for the quantitative comparison of XAI methods. A method for selecting an appropriate perturbing value is also proposed based on the concept of information entropy to yield reliable perturbation analysis results. For the experiment, a simple nuclear power plant (NPP) accident diagnosis model was developed to reflect the characteristics of the nuclear domain, and four XAI methods were applied for comparative analysis. The experimental results demonstrate that perturbation analysis and the proposed method are effective for quantitatively comparing the performance of XAI methods.
{"title":"Quantitative comparison of explainable artificial intelligence methods for nuclear power plant accident diagnosis models","authors":"Seung Geun Kim , Seunghyoung Ryu , Kyungho Jin , Hyeonmin Kim","doi":"10.1016/j.pnucene.2025.105605","DOIUrl":"10.1016/j.pnucene.2025.105605","url":null,"abstract":"<div><div>The rapid advancement of artificial intelligence (AI) technology based on deep neural networks (DNNs) has spurred active development of DNN-based models in the nuclear domain. Due to the black-box nature of these models and the issue of low explainability, their practical application in safety-critical domains is hindered. To address this, numerous explainable AI (XAI) methods have been proposed. However, the selection of an appropriate XAI method is crucial as its performance significantly depends on various factors; nonetheless, comparative studies of XAI methods are limited within the nuclear domain. This study employs perturbation analysis for the quantitative comparison of XAI methods. A method for selecting an appropriate perturbing value is also proposed based on the concept of information entropy to yield reliable perturbation analysis results. For the experiment, a simple nuclear power plant (NPP) accident diagnosis model was developed to reflect the characteristics of the nuclear domain, and four XAI methods were applied for comparative analysis. The experimental results demonstrate that perturbation analysis and the proposed method are effective for quantitatively comparing the performance of XAI methods.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105605"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141793","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}
Isothermal steam oxidation of Indian pressurized heavy water reactors (IPHWRs) cladding (Zircaloy-4) has been studied in the temperature range of 1000–1500 °C. Growth kinetics of inherently brittle phases, oxide and oxygen stabilized α-Zr, developed during the high temperature steam oxidation were established. A ductile phase known as prior β-Zr, lying either underneath the α-Zr(O), in case of a single sided oxidation, or in between two layers of α-Zr(O) at the inner and the outer surface of the clad, governs the extent of embrittlement of the clad. Thickness of prior β-Zr is decided by the rate of growth of oxide and oxygen stabilized α-Zr. The kinetic rate constant Kp (cm2/sec) obeyed a parabolic rate equation. The Arrhenius expression of parabolic rate constants for oxide scale and α-Zr(O) layer growth compares well with those established by other investigators. An in-house oxidation model, OXYCON was evaluated against the experimental data of oxide and α-Zr(O) layer thickness. The model under-predicted the oxide layer thickness and over-predicted the α-Zr(O) layer thickness in its existing form. However the prediction of oxide, α-Zr (O) and total oxide+α-Zr(O) improved to a significant extent by using kinetic equations derived in the present study done on the indigenously fabricated Zircaloy-4 cladding. Hence, successively the prediction of the prior β-Zr layer thickness could also be done with further accuracy.
{"title":"Study of oxide and α-Zr(O) growth kinetics in the temperature range 1000–1500 °C during steam oxidation of Indian PHWR cladding","authors":"Tapan K. Sawarn , Suparna Banerjee , Saee Jagtap , Alexander Rajath , Raman Saini , Sayandeep Kundu , Sripooja Mishra , P.P. Nanekar","doi":"10.1016/j.pnucene.2025.105632","DOIUrl":"10.1016/j.pnucene.2025.105632","url":null,"abstract":"<div><div>Isothermal steam oxidation of Indian pressurized heavy water reactors (IPHWRs) cladding (Zircaloy-4) has been studied in the temperature range of 1000–1500 °C. Growth kinetics of inherently brittle phases, oxide and oxygen stabilized α-Zr, developed during the high temperature steam oxidation were established. A ductile phase known as prior β-Zr, lying either underneath the α-Zr(O), in case of a single sided oxidation, or in between two layers of α-Zr(O) at the inner and the outer surface of the clad, governs the extent of embrittlement of the clad. Thickness of prior β-Zr is decided by the rate of growth of oxide and oxygen stabilized α-Zr. The kinetic rate constant K<sub>p</sub> (cm<sup>2</sup>/sec) obeyed a parabolic rate equation. The Arrhenius expression of parabolic rate constants for oxide scale and α-Zr(O) layer growth compares well with those established by other investigators. An in-house oxidation model, OXYCON was evaluated against the experimental data of oxide and α-Zr(O) layer thickness. The model under-predicted the oxide layer thickness and over-predicted the α-Zr(O) layer thickness in its existing form. However the prediction of oxide, α-Zr (O) and total oxide+α-Zr(O) improved to a significant extent by using kinetic equations derived in the present study done on the indigenously fabricated Zircaloy-4 cladding. Hence, successively the prediction of the prior β-Zr layer thickness could also be done with further accuracy.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105632"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143140931","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}
During a loss-of-coolant accident in a pressurized water reactor (PWR), there is a risk that pressurized thermal shock (PTS) may occur due to the rapid cooling of the downcomer wall caused by the emergency core cooling (ECC) water injected into the cold leg that flows into the downcomer. To predict the PTS, it is necessary to accurately predict the temperature of the ECC water, the collision of the water jet on the downcomer wall, the velocity, thickness, and the spread of the liquid film. Therefore, to help understand this flow phenomenon, we reviewed studies on free outflow from a circular pipe. In many previous studies, experimental findings on the flow regimes, transition conditions between flow regimes, characteristics of the flow shape have been obtained in a form that was almost consistent with each other. And, through theoretical analysis, it was also possible to predict the flow regime and flow surface shape to a certain extent quantitatively. In contrast, when considering the flow from the cold leg to the downcomer, it is necessary to deal with the flow field in a specific situation, such as the flow into a narrow gap rather than a free space, the existence of rounded corners at the outlet of the circular pipe, and the influence of steam flow flowing from the core to the cold leg. However, few previous studies consider these factors, so we summarized them as knowledge that needs to be accumulated in the future. In this review article, 30 references are included.
{"title":"Free outflow from the end of a horizontal circular pipe related to flow from the PWR cold leg to the downcomer","authors":"Akira Satou , Takashi Hibiki , Ryo Ikeda , Yasuteru Sibamoto","doi":"10.1016/j.pnucene.2024.105593","DOIUrl":"10.1016/j.pnucene.2024.105593","url":null,"abstract":"<div><div>During a loss-of-coolant accident in a pressurized water reactor (PWR), there is a risk that pressurized thermal shock (PTS) may occur due to the rapid cooling of the downcomer wall caused by the emergency core cooling (ECC) water injected into the cold leg that flows into the downcomer. To predict the PTS, it is necessary to accurately predict the temperature of the ECC water, the collision of the water jet on the downcomer wall, the velocity, thickness, and the spread of the liquid film. Therefore, to help understand this flow phenomenon, we reviewed studies on free outflow from a circular pipe. In many previous studies, experimental findings on the flow regimes, transition conditions between flow regimes, characteristics of the flow shape have been obtained in a form that was almost consistent with each other. And, through theoretical analysis, it was also possible to predict the flow regime and flow surface shape to a certain extent quantitatively. In contrast, when considering the flow from the cold leg to the downcomer, it is necessary to deal with the flow field in a specific situation, such as the flow into a narrow gap rather than a free space, the existence of rounded corners at the outlet of the circular pipe, and the influence of steam flow flowing from the core to the cold leg. However, few previous studies consider these factors, so we summarized them as knowledge that needs to be accumulated in the future. In this review article, 30 references are included.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105593"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141577","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 : 2025-02-01DOI: 10.1016/j.pnucene.2024.105589
Hongjian Zhang, Qing Zhu, Haiyan Xiao, Liguo Zhang, Tao Ma
Pebble-bed high-temperature gas-cooled reactors operate under a continuous refueling regime, rendering the trajectory of individual fuel spheres within the reactor core indeterminate. This inherent lack of tracking capability results in an obscured burnup history for the fuel spheres. Consequently, the conventional approach of deducing the nuclide inventory in the nuclear fuel of pebble-bed reactors through burnup calculations, predicated on initial fuel composition, is infeasible. Presently, while the online Burnup Measurement System integrated with High Temperature Reactor-Pebble bed Modules allows for the acquisition of gamma-ray spectra, enabling the measurement of certain nuclide activities, the direct non-destructive measurement of fissile isotopes remains elusive. This limitation poses a significant challenge to the field of nuclear material accounting, necessitating innovative methodologies for accurate inventory assessment.
This study presents a comprehensive approach to enhance the predictive accuracy of fissile isotopes and Zr-95 in pebble-bed reactors, encompassing three research facets. Initially, the Discrete Element Method (DEM) model is employed to simulate the core of a pebble-bed reactor. The model is calibrated against three-dimensional pebble flow experimental data by selecting an appropriate contact model and tuning model parameters to ensure consistency in flow dispersion, radial velocity ratio, and other key metrics. Subsequently, irradiation histories for the fuel spheres are constructed based on their trajectories within the reactor core. This data is then integrated with a nuclide inventory calculation software to simulate burnup information for the fuel spheres. Finally, the study employs multivariate analysis techniques, including Principal Component Analysis (PCA), ridge regression, and the random forest model, to predict fissile isotopes and Zr-95 from activity of online-measurable radionuclides. The predictive accuracy of our approach is appraised by correlating the outcomes yielded by a multivariate linear model, trained with simulated datasets, against those derived from experimental data.
This paper establishes a computational method for fissile isotopes in pebble-bed high-temperature reactors, enhancing the accuracy of nuclear material accounting, with a prediction error of only 4.9% for Zr-95, which is the most accurate experimental data that can be obtained. It demonstrates that online nuclide content contains valuable information worth further exploration and validates the applicability of multivariate analysis methods in nuclear material calculations.
{"title":"Multivariate analysis quantifying amount of fissile isotopes for pebble-bed high temperature reactors","authors":"Hongjian Zhang, Qing Zhu, Haiyan Xiao, Liguo Zhang, Tao Ma","doi":"10.1016/j.pnucene.2024.105589","DOIUrl":"10.1016/j.pnucene.2024.105589","url":null,"abstract":"<div><div>Pebble-bed high-temperature gas-cooled reactors operate under a continuous refueling regime, rendering the trajectory of individual fuel spheres within the reactor core indeterminate. This inherent lack of tracking capability results in an obscured burnup history for the fuel spheres. Consequently, the conventional approach of deducing the nuclide inventory in the nuclear fuel of pebble-bed reactors through burnup calculations, predicated on initial fuel composition, is infeasible. Presently, while the online Burnup Measurement System integrated with High Temperature Reactor-Pebble bed Modules allows for the acquisition of gamma-ray spectra, enabling the measurement of certain nuclide activities, the direct non-destructive measurement of fissile isotopes remains elusive. This limitation poses a significant challenge to the field of nuclear material accounting, necessitating innovative methodologies for accurate inventory assessment.</div><div>This study presents a comprehensive approach to enhance the predictive accuracy of fissile isotopes and Zr-95 in pebble-bed reactors, encompassing three research facets. Initially, the Discrete Element Method (DEM) model is employed to simulate the core of a pebble-bed reactor. The model is calibrated against three-dimensional pebble flow experimental data by selecting an appropriate contact model and tuning model parameters to ensure consistency in flow dispersion, radial velocity ratio, and other key metrics. Subsequently, irradiation histories for the fuel spheres are constructed based on their trajectories within the reactor core. This data is then integrated with a nuclide inventory calculation software to simulate burnup information for the fuel spheres. Finally, the study employs multivariate analysis techniques, including Principal Component Analysis (PCA), ridge regression, and the random forest model, to predict fissile isotopes and Zr-95 from activity of online-measurable radionuclides. The predictive accuracy of our approach is appraised by correlating the outcomes yielded by a multivariate linear model, trained with simulated datasets, against those derived from experimental data.</div><div>This paper establishes a computational method for fissile isotopes in pebble-bed high-temperature reactors, enhancing the accuracy of nuclear material accounting, with a prediction error of only 4.9% for Zr-95, which is the most accurate experimental data that can be obtained. It demonstrates that online nuclide content contains valuable information worth further exploration and validates the applicability of multivariate analysis methods in nuclear material calculations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105589"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141588","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 : 2025-02-01DOI: 10.1016/j.pnucene.2025.105603
Yan Zhou, Sixu Yan, Ao Gao, Zhenhua Xiong
Radiation field reconstruction of nuclear facilities can help to visualize the ionizing radiation distribution and to ensure operator safety, which is of great importance in the radiation protection field. Currently, radiation sources in nuclear facilities are strictly controlled, and their locations and radioactivities are clearly known. However, the environment and the devices surrounding the sources are usually so complex that their shielding effect on radiation rays cannot be easily acquired. This results in a challenging task of the radiation field acquirement. Many field reconstruction methods relying on precise radiation attenuation models of the environment will fail in this situation. Aiming at the problem above, this paper proposes an equivalent radiation attenuation model (ERAM), which computes the equalized shielding effect of the complex environment with limited detection data, and then reconstructs the field. Mobile robot can be applied to obtain necessary radiation data in the scene, providing a complete and feasible strategy for constructing the ERAM and the radiation field. In the simulation, this paper demonstrates the universality of the proposed method with different sources and environment situations. In the experiment, mobile robot equipped with radiation sensors is applied in two nuclear facilities, and the radiation field is successfully reconstructed. The results show that the proposed method provides an effective and convenient solution for the radiation field reconstruction.
{"title":"Equivalent radiation attenuation method for nuclear radiation field reconstruction","authors":"Yan Zhou, Sixu Yan, Ao Gao, Zhenhua Xiong","doi":"10.1016/j.pnucene.2025.105603","DOIUrl":"10.1016/j.pnucene.2025.105603","url":null,"abstract":"<div><div>Radiation field reconstruction of nuclear facilities can help to visualize the ionizing radiation distribution and to ensure operator safety, which is of great importance in the radiation protection field. Currently, radiation sources in nuclear facilities are strictly controlled, and their locations and radioactivities are clearly known. However, the environment and the devices surrounding the sources are usually so complex that their shielding effect on radiation rays cannot be easily acquired. This results in a challenging task of the radiation field acquirement. Many field reconstruction methods relying on precise radiation attenuation models of the environment will fail in this situation. Aiming at the problem above, this paper proposes an equivalent radiation attenuation model (ERAM), which computes the equalized shielding effect of the complex environment with limited detection data, and then reconstructs the field. Mobile robot can be applied to obtain necessary radiation data in the scene, providing a complete and feasible strategy for constructing the ERAM and the radiation field. In the simulation, this paper demonstrates the universality of the proposed method with different sources and environment situations. In the experiment, mobile robot equipped with radiation sensors is applied in two nuclear facilities, and the radiation field is successfully reconstructed. The results show that the proposed method provides an effective and convenient solution for the radiation field reconstruction.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105603"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141805","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}
This paper centers on the intricate phenomenon of water hammer observed in steam pipelines during nuclear power plant diagnostics. Utilizing computational fluid dynamics (CFD), the paper simulates the transient response of condensate-induced water hammer (CIWH) behavior. A sophisticated phase transition model is applied with the goal of precisely delineating the phase change process and its subsequent impact on system dynamics. For purposes of investigating the influence of atmospheric vent valves on CIWH, the study introduces a resistance term in the simulation model to embody the dynamic valve control, thereby comprehensively examining its interplay with CIWH phenomena. The correspondence between simulation findings and actual steam pipeline operations is carefully compared for verification. Moreover, this paper explores the influence of key parameters on CIWH and evaluates the strength of CIWH by using Froude (Fr) and modified Jakob (Ja) numbers. Within a certain range, it is found that smaller Fr and Ja result in segmental plug flow in the tube and the strength of CIWH increases with increasing Fr and Ja. However, larger Fr and Ja cause the flow pattern in the tube to change and the intensity of CIWH also changes. We have analyzed in depth the effect of different thermodynamic parameters on the mass transfer rate during CIWH and made an assessment in conjunction with the dynamics of the valves. Ultimately, the numerical simulation method established in this paper effectively reproduces the process of complex water hammer and lays the foundation for subsequent model optimization and triggering mechanism studies.
{"title":"Numerical simulation of condensate-induced water hammer with dynamic valve control using an advanced phase change model","authors":"Zihao Zhang , Jiaming Zhao , Danfeng Zhao , Ruiyang Tu , Feng Xiong , Zhengyu Chen , Wentao Guo , Shengfei Wang","doi":"10.1016/j.pnucene.2025.105631","DOIUrl":"10.1016/j.pnucene.2025.105631","url":null,"abstract":"<div><div>This paper centers on the intricate phenomenon of water hammer observed in steam pipelines during nuclear power plant diagnostics. Utilizing computational fluid dynamics (CFD), the paper simulates the transient response of condensate-induced water hammer (CIWH) behavior. A sophisticated phase transition model is applied with the goal of precisely delineating the phase change process and its subsequent impact on system dynamics. For purposes of investigating the influence of atmospheric vent valves on CIWH, the study introduces a resistance term in the simulation model to embody the dynamic valve control, thereby comprehensively examining its interplay with CIWH phenomena. The correspondence between simulation findings and actual steam pipeline operations is carefully compared for verification. Moreover, this paper explores the influence of key parameters on CIWH and evaluates the strength of CIWH by using Froude (Fr) and modified Jakob (Ja) numbers. Within a certain range, it is found that smaller Fr and Ja result in segmental plug flow in the tube and the strength of CIWH increases with increasing Fr and Ja. However, larger Fr and Ja cause the flow pattern in the tube to change and the intensity of CIWH also changes. We have analyzed in depth the effect of different thermodynamic parameters on the mass transfer rate during CIWH and made an assessment in conjunction with the dynamics of the valves. Ultimately, the numerical simulation method established in this paper effectively reproduces the process of complex water hammer and lays the foundation for subsequent model optimization and triggering mechanism studies.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105631"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143140930","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 : 2025-02-01DOI: 10.1016/j.pnucene.2024.105587
Yue Ma , Qianfeng Liu , Yunjie Jiao , Zhu Cui , Debo Yang , Dongyu Wang , Xuanjiang Dong , Yongchang Zhu , Li Junfeng , Wang Jianlong
The safe disposal of high-level nuclear radioactive waste has become an increasingly concerned problem. Glass solidification by Joule-heated ceramic melter is the mainly used technology for vitrification of high-level radioactive waste at present. In this paper, the temperature field and velocity field of molten glass flow in the furnace during the glass solidification process, and the heat and mass transfer phenomenon in the ceramic melter was analyzed and simulated using the computational fluid dynamics software ANSYS Fluent. The numerical simulation of precious metal deposition was also carried out to obtain the volume fraction distribution of precious metal inside the glass. The results showed that near the entrance and near the exit, the direction of glass flow changed, and there is a concentration of heavy metals. Near the lower furnace walls, there is a slight metal disposition phenomenon. In the cold core, the temperature of the glass near the wall can reach 1405 K and the velocity can reach 0.05 m/s.
{"title":"Numerical simulation of flow field and metal deposition of Joule-heated ceramic melter for vitrification of radioactive waste","authors":"Yue Ma , Qianfeng Liu , Yunjie Jiao , Zhu Cui , Debo Yang , Dongyu Wang , Xuanjiang Dong , Yongchang Zhu , Li Junfeng , Wang Jianlong","doi":"10.1016/j.pnucene.2024.105587","DOIUrl":"10.1016/j.pnucene.2024.105587","url":null,"abstract":"<div><div>The safe disposal of high-level nuclear radioactive waste has become an increasingly concerned problem. Glass solidification by Joule-heated ceramic melter is the mainly used technology for vitrification of high-level radioactive waste at present. In this paper, the temperature field and velocity field of molten glass flow in the furnace during the glass solidification process, and the heat and mass transfer phenomenon in the ceramic melter was analyzed and simulated using the computational fluid dynamics software ANSYS Fluent. The numerical simulation of precious metal deposition was also carried out to obtain the volume fraction distribution of precious metal inside the glass. The results showed that near the entrance and near the exit, the direction of glass flow changed, and there is a concentration of heavy metals. Near the lower furnace walls, there is a slight metal disposition phenomenon. In the cold core, the temperature of the glass near the wall can reach 1405 K and the velocity can reach 0.05 m/s.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105587"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141635","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 : 2025-02-01DOI: 10.1016/j.pnucene.2024.105549
Peizheng Hu, Lili Tong, Xuewu Cao
The aerosol suspended in containment can be removed by the spray system, mitigating the potential risk of radioactive release during severe accidents. To analyze aerosol removal efficiency due to mechanical and phoretic effects of spray droplets, experiments were conducted under various thermal-hydraulic conditions at the Containment Aerosol and Thermal-Hydraulics (CATH) facility. Initial thermal-hydraulic conditions involved either pure air or gas mixture, with a pressure of 5 bar(a), steam fractions of 40%vol and 70%vol, and a temperature range of 20 °C–145 °C, simulating the severe accident conditions. The sprays were produced by a hollow cone nozzle with a volume mean diameter (VMD) of 370 μm at a constant spray flow rate. Results indicate that the removal efficiencies of mechanical effects, thermophoresis, and diffusiophoresis are comparable at a steam fraction of 40%vol, while diffusiophoretic effects increase to three times when steam fraction reaches 70%vol. Additionally, the evaluations of the present aerosol removal models show significant deviations from experiments, with inertial impaction is highly sensitive to increases in particle size, while the phoretic mechanism underestimates the effect of steam condensation.
{"title":"Investigation on aerosol removal by spray droplets under severe accident conditions","authors":"Peizheng Hu, Lili Tong, Xuewu Cao","doi":"10.1016/j.pnucene.2024.105549","DOIUrl":"10.1016/j.pnucene.2024.105549","url":null,"abstract":"<div><div>The aerosol suspended in containment can be removed by the spray system, mitigating the potential risk of radioactive release during severe accidents. To analyze aerosol removal efficiency due to mechanical and phoretic effects of spray droplets, experiments were conducted under various thermal-hydraulic conditions at the Containment Aerosol and Thermal-Hydraulics (CATH) facility. Initial thermal-hydraulic conditions involved either pure air or gas mixture, with a pressure of 5 bar(a), steam fractions of 40%vol and 70%vol, and a temperature range of 20 °C–145 °C, simulating the severe accident conditions. The sprays were produced by a hollow cone nozzle with a volume mean diameter (VMD) of 370 μm at a constant spray flow rate. Results indicate that the removal efficiencies of mechanical effects, thermophoresis, and diffusiophoresis are comparable at a steam fraction of 40%vol, while diffusiophoretic effects increase to three times when steam fraction reaches 70%vol. Additionally, the evaluations of the present aerosol removal models show significant deviations from experiments, with inertial impaction is highly sensitive to increases in particle size, while the phoretic mechanism underestimates the effect of steam condensation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105549"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141678","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 : 2025-02-01DOI: 10.1016/j.pnucene.2024.105539
Hanbee Na, Namchul Cho
Using the CFD-DEM method, numerical calculations are conducted to analyze the shape of particle beds formed in two small-sized pre-flooded reactor cavities, aiming to understand the effect of the reactor cavity's geometry and the reactor vessel's relative position on the shape of the debris beds. The first reactor cavity floor is circular, with the reactor vessel positioned at the center. The second reactor cavity resembles a corridor, with the reactor vessel located near the end of the corridor. In both cases, mound-shaped particle beds are formed. However, the particle bed in the second reactor cavity is densely stacked in a narrower region, with a large part of the side surface blocked by wall. As a result of this study, it is expected that the radioactive decay heat in the pre-flooded corridor-type reactor cavity with an off-centered reactor vessel is more concentrated than in the circular reactor cavity with a centered reactor vessel.
{"title":"A comparative study on the effect of reactor cavity shape and location of reactor vessel on the shape of particle bed formed in pre-flooded reactor cavity using the CFD-DEM method","authors":"Hanbee Na, Namchul Cho","doi":"10.1016/j.pnucene.2024.105539","DOIUrl":"10.1016/j.pnucene.2024.105539","url":null,"abstract":"<div><div>Using the CFD-DEM method, numerical calculations are conducted to analyze the shape of particle beds formed in two small-sized pre-flooded reactor cavities, aiming to understand the effect of the reactor cavity's geometry and the reactor vessel's relative position on the shape of the debris beds. The first reactor cavity floor is circular, with the reactor vessel positioned at the center. The second reactor cavity resembles a corridor, with the reactor vessel located near the end of the corridor. In both cases, mound-shaped particle beds are formed. However, the particle bed in the second reactor cavity is densely stacked in a narrower region, with a large part of the side surface blocked by wall. As a result of this study, it is expected that the radioactive decay heat in the pre-flooded corridor-type reactor cavity with an off-centered reactor vessel is more concentrated than in the circular reactor cavity with a centered reactor vessel.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105539"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141681","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 : 2025-02-01DOI: 10.1016/j.pnucene.2025.105611
Shuang-qing Chen , Sheng-hui Liu , Dian-le Wang , Hai-jie Wu , Ruo-han Zheng , Min-yun Liu , Hou-jun Gong , Jia-tao Meng , Yan-ping Huang , Xiao-liang Zhu
To deep understanding the heat transfer and flow process in zigzag-type printed circuit heat exchanger (PCHE) for advanced nuclear system, the coupling effects of bend corners are studied numerically. Bend corner, as the fundamental unit of zigzag channel, has great influence on its heat transfer and hydraulic performance of zigzag-type PCHE. In this study, the effects of bend corner and the coupling effects of adjacent bend corners were investigated, with bending angle 10°–40°, and Reynolds number (Re) 2500–20000. Concept of Influence Distance (ID) was proposed to describe the effects of bend corner on the convective heat transfer coefficient (HTC) and Fanning friction factor (f) quantificationally. It is found that the presence of the bend corner leads to secondary flow, recirculation and boundary layer separation. When bending angle is about 17°, ID-f almost keeps as a constant value, about 21 mm, for Re 2500–20000. And each curve of ID-HTC reaches a peak value when bending angle reaches about 30°. What's more, the coupling effects of adjacent bend corners of zigzag channels was demonstrated by comparing the heat transfer and hydraulic characteristics between bend corners arranged alternatingly and non-alternatingly. The shorter of midstream length and the higher Re, the stronger coupling effects of the adjacent bend corners. For the cases with Re = 20000, bending angle 30° and midstream length 10 mm–30 mm, the HTC near the second bend corner is about 9% lower than that of the first bend corner. The coupling effects influences the heat transfer and hydraulic characteristics obviously.
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