Pub Date : 2026-02-04DOI: 10.1016/j.pnucene.2026.106291
Jia Tang , Yin Hu , Yi Xia , Jixue Sui , Jiajia Yang , Yunming Chen , Xiaogen Xiong , Yi Wang , Qi Cao
He-Xe binary mixture is a leading candidate for the fourth-generation gas-cooled reactor coolants. this study utilized MJTR to conduct in-pile irradiation experiments on quartz target components filled with Xe gas. The cumulative neutron fluence exceeded 1.0E16 n/cm2. And a qualitative to semi-quantitative relationship between Xe impurities yield and neutron fluence was established. Based on the experimental results, the radioactivity derived from neutron irradiation of Xe in specific reactor system was estimated. The chemistry and deposition processes of these impurities was predicted. The results show that CsI and Cs2Te are thermodynamically formable. Once formed, they are in solid state and unlikely to spontaneously decompose into atoms or ions. At temperatures below 1300–1400 K and 1100–1200 K, CsI and Cs2Te may spontaneously undergo polymerization reactions to form dimers, which further crystallize, leading to the formation of macroscopic crystals. The main impurity Te would react with Inconel 617, a promising structure material, to form NiTe0.69. This study deepens the understanding of the physical and chemical behavior of Xe under neutron irradiation and provides valuable guidance for the design and operational strategies of He-Xe coolant-based gas-cooled reactor systems.
{"title":"Analysis of key nuclides in Xe irradiated by neutrons and their chemical behavior","authors":"Jia Tang , Yin Hu , Yi Xia , Jixue Sui , Jiajia Yang , Yunming Chen , Xiaogen Xiong , Yi Wang , Qi Cao","doi":"10.1016/j.pnucene.2026.106291","DOIUrl":"10.1016/j.pnucene.2026.106291","url":null,"abstract":"<div><div>He-Xe binary mixture is a leading candidate for the fourth-generation gas-cooled reactor coolants. this study utilized MJTR to conduct in-pile irradiation experiments on quartz target components filled with Xe gas. The cumulative neutron fluence exceeded 1.0E16 n/cm<sup>2</sup>. And a qualitative to semi-quantitative relationship between Xe impurities yield and neutron fluence was established. Based on the experimental results, the radioactivity derived from neutron irradiation of Xe in specific reactor system was estimated. The chemistry and deposition processes of these impurities was predicted. The results show that CsI and Cs<sub>2</sub>Te are thermodynamically formable. Once formed, they are in solid state and unlikely to spontaneously decompose into atoms or ions. At temperatures below 1300–1400 K and 1100–1200 K, CsI and Cs<sub>2</sub>Te may spontaneously undergo polymerization reactions to form dimers, which further crystallize, leading to the formation of macroscopic crystals. The main impurity Te would react with Inconel 617, a promising structure material, to form NiTe<sub>0.69</sub>. This study deepens the understanding of the physical and chemical behavior of Xe under neutron irradiation and provides valuable guidance for the design and operational strategies of He-Xe coolant-based gas-cooled reactor systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106291"},"PeriodicalIF":3.2,"publicationDate":"2026-02-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189712","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 : 2026-02-03DOI: 10.1016/j.pnucene.2026.106287
Zuolong Wu , Jiejin Cai , Facheng Chen , Heng Xie
In order to study the influence of the spacer grid on thermal-hydraulic characteristics in a 4 × 4 rod bundle channel, a subcooled boiling multiphase flow model has been established, and the reliability of the numerical model has been verified. The calculated and experimental volume fraction of vapor and CHF are compared and are in good agreement. The CFD software STAR-CCM+ is used to study the influence of the spacer grid on the thermal-hydraulic characteristics, and to obtain the CHF and distributions of flow field in the rod bundle channel under multiple operating conditions. The results show that the spacer grid induces a CHF enhancement of more than 5.6%, an increase of the average secondary flow velocity maximum by more than 23 times, a reduction of maximum volume of vapor by more than 35.2%, and an increase in the total pressure drop in the rod bundle channel by more than 0.7 times. The dimple and spring structure dominate the pressure drop contribution, while the mixing vane structure performs better in secondary flow enhancement and lateral churning.
{"title":"Simulation study of the influence of the spacer grid on thermal-hydraulic characteristics in the rod bundle channel","authors":"Zuolong Wu , Jiejin Cai , Facheng Chen , Heng Xie","doi":"10.1016/j.pnucene.2026.106287","DOIUrl":"10.1016/j.pnucene.2026.106287","url":null,"abstract":"<div><div>In order to study the influence of the spacer grid on thermal-hydraulic characteristics in a 4 × 4 rod bundle channel, a subcooled boiling multiphase flow model has been established, and the reliability of the numerical model has been verified. The calculated and experimental volume fraction of vapor and CHF are compared and are in good agreement. The CFD software STAR-CCM+ is used to study the influence of the spacer grid on the thermal-hydraulic characteristics, and to obtain the CHF and distributions of flow field in the rod bundle channel under multiple operating conditions. The results show that the spacer grid induces a CHF enhancement of more than 5.6%, an increase of the average secondary flow velocity maximum by more than 23 times, a reduction of maximum volume of vapor by more than 35.2%, and an increase in the total pressure drop in the rod bundle channel by more than 0.7 times. The dimple and spring structure dominate the pressure drop contribution, while the mixing vane structure performs better in secondary flow enhancement and lateral churning.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106287"},"PeriodicalIF":3.2,"publicationDate":"2026-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189700","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 : 2026-02-03DOI: 10.1016/j.pnucene.2026.106288
A. Milena-Pérez, F. Feria, C. Aguado, S. Fernández-Carretero, L. Gutiérrez, N. Rodríguez-Villagra, H. Galán, L.E. Herranz
In this work, we have studied the oxidation behavior of a fresh UO2 surrogate in representative conditions of a dry interim storage for SNF, both in terms of temperature (200–400 °C) and oxygen partial pressure (0.1–21% O2). The reaction has been studied in-situ by TGA and ex-situ by autoclaves. XRD and Rietveld's refinement has been used for quantification of uranium oxidized phases, with particular interest in U3O8. The predominance of temperature over oxygen concentration in the formation of U3O8 has been confirmed. TGA results have shown that, at 200 and 250 °C, no U3O8 has been detected after 10 h of thermal treatment, not even in air (21% O2). Only at 275 °C, U3O8 has been detected at 21 and 1% O2. This result allows better defining the temperature threshold of U3O8 formation in unirradiated UO2. On the upper limit, 350 and 400 °C, the reaction proceeds up to quantitative formation (i.e., more than 90%) of U3O8, even with very low oxygen content (as low as 0.1% O2). Finally, at intermediate temperatures (i.e., 300 °C), oxygen partial pressure starts playing a role, being the extent of the oxidation lower with lower oxygen concentration. This effect has been seen when studying the reaction at longer times by using ex-situ autoclaves. However, given enough time, all the fuels end showing a majority of U3O8, which is between 6 and 24 h for 10 and 1% O2 and between 24 h and 1 week for 0.1% O2. In all the experiments, 21 and 10% O2 behave in a very similar way for all the temperatures, but there is a substantial change when decreasing oxygen to 1% O2, being in this case the oxidation much more limited. This could be an indication of an oxygen concentration threshold that could trigger the nucleation and growth of U3O8 in a greater extent. Based on the experimental work and previous modelling studies, an analytical methodology to assess the cladding integrity during UO2 oxidation has been developed by adapting FRAPCON-xt, a home-extended fuel performance code to dry storage.
{"title":"UO2 oxidation in dry storage conditions: from new data to exploratory modelling of consequences","authors":"A. Milena-Pérez, F. Feria, C. Aguado, S. Fernández-Carretero, L. Gutiérrez, N. Rodríguez-Villagra, H. Galán, L.E. Herranz","doi":"10.1016/j.pnucene.2026.106288","DOIUrl":"10.1016/j.pnucene.2026.106288","url":null,"abstract":"<div><div>In this work, we have studied the oxidation behavior of a fresh UO<sub>2</sub> surrogate in representative conditions of a dry interim storage for SNF, both in terms of temperature (200–400 °C) and oxygen partial pressure (0.1–21% O<sub>2</sub>). The reaction has been studied <em>in-situ</em> by TGA and <em>ex-situ</em> by autoclaves. XRD and Rietveld's refinement has been used for quantification of uranium oxidized phases, with particular interest in U<sub>3</sub>O<sub>8</sub>. The predominance of temperature over oxygen concentration in the formation of U<sub>3</sub>O<sub>8</sub> has been confirmed. TGA results have shown that, at 200 and 250 °C, no U<sub>3</sub>O<sub>8</sub> has been detected after 10 h of thermal treatment, not even in air (21% O<sub>2</sub>). Only at 275 °C, U<sub>3</sub>O<sub>8</sub> has been detected at 21 and 1% O<sub>2</sub>. This result allows better defining the temperature threshold of U<sub>3</sub>O<sub>8</sub> formation in unirradiated UO<sub>2</sub>. On the upper limit, 350 and 400 °C, the reaction proceeds up to quantitative formation (<em>i.e.,</em> more than 90%) of U<sub>3</sub>O<sub>8</sub>, even with very low oxygen content (as low as 0.1% O<sub>2</sub>). Finally, at intermediate temperatures (<em>i.e.,</em> 300 °C), oxygen partial pressure starts playing a role, being the extent of the oxidation lower with lower oxygen concentration. This effect has been seen when studying the reaction at longer times by using <em>ex-situ</em> autoclaves. However, given enough time, all the fuels end showing a majority of U<sub>3</sub>O<sub>8</sub>, which is between 6 and 24 h for 10 and 1% O<sub>2</sub> and between 24 h and 1 week for 0.1% O<sub>2</sub>. In all the experiments, 21 and 10% O<sub>2</sub> behave in a very similar way for all the temperatures, but there is a substantial change when decreasing oxygen to 1% O<sub>2</sub>, being in this case the oxidation much more limited. This could be an indication of an oxygen concentration threshold that could trigger the nucleation and growth of U<sub>3</sub>O<sub>8</sub> in a greater extent. Based on the experimental work and previous modelling studies, an analytical methodology to assess the cladding integrity during UO<sub>2</sub> oxidation has been developed by adapting FRAPCON-xt, a home-extended fuel performance code to dry storage.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106288"},"PeriodicalIF":3.2,"publicationDate":"2026-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189226","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 : 2026-02-03DOI: 10.1016/j.pnucene.2026.106285
Xiaozhong Wang , Qi Sun , Wei Peng , Yinhai Zhu , Peixue Jiang , Suyuan Yu
Resuspension significantly influences the transport and deposition of graphite particles within the turbine of high-temperature gas-cooled reactor (HTGR). This work involved setting up a visualized experimental setup to observe particle resuspension under film cooling conditions. High-speed cameras and hot-wire anemometers were employed to measure the dynamic process of particle resuspension and the characteristics of the near-wall flow field. Based on the torque balance model, a particle resuspension model suitable for the HTGR turbine conditions was developed, and the accuracy of the model was verified with experimental results. Combining experimental measurements and numerical simulations, the mechanism by which film cooling affects particle resuspension was investigated, and the dominant role of counter-rotating vortices was revealed. Furthermore, the influence of key factors such as flow velocity, blowing ratio, and particle size on resuspension characteristics was systematically analyzed. Finally, the particle deposition process and resuspension process were coupled for analysis, and the distribution characteristics of particle deposition on the turbine surface under film cooling conditions were studied. The findings from this study not only offer experimental insights into particle resuspension behavior but also contribute valuable guidance for the design and safety assessment of HTGR helium turbines.
{"title":"Experimental and numerical study of particle resuspension on turbine surface with film cooling","authors":"Xiaozhong Wang , Qi Sun , Wei Peng , Yinhai Zhu , Peixue Jiang , Suyuan Yu","doi":"10.1016/j.pnucene.2026.106285","DOIUrl":"10.1016/j.pnucene.2026.106285","url":null,"abstract":"<div><div>Resuspension significantly influences the transport and deposition of graphite particles within the turbine of high-temperature gas-cooled reactor (HTGR). This work involved setting up a visualized experimental setup to observe particle resuspension under film cooling conditions. High-speed cameras and hot-wire anemometers were employed to measure the dynamic process of particle resuspension and the characteristics of the near-wall flow field. Based on the torque balance model, a particle resuspension model suitable for the HTGR turbine conditions was developed, and the accuracy of the model was verified with experimental results. Combining experimental measurements and numerical simulations, the mechanism by which film cooling affects particle resuspension was investigated, and the dominant role of counter-rotating vortices was revealed. Furthermore, the influence of key factors such as flow velocity, blowing ratio, and particle size on resuspension characteristics was systematically analyzed. Finally, the particle deposition process and resuspension process were coupled for analysis, and the distribution characteristics of particle deposition on the turbine surface under film cooling conditions were studied. The findings from this study not only offer experimental insights into particle resuspension behavior but also contribute valuable guidance for the design and safety assessment of HTGR helium turbines.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106285"},"PeriodicalIF":3.2,"publicationDate":"2026-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189704","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 : 2026-02-02DOI: 10.1016/j.pnucene.2026.106271
Shaoting Jia, Takashi Hibiki
The two-fluid model is widely employed in thermal-hydraulic simulations of gas-liquid two-phase flows. Its predictive capability is strongly dependent on the accurate estimation of the interfacial area concentration (IAC), a key parameter governing the transfer of mass, momentum, and energy across the interphase. In dispersed flows, bubbles vary widely in size and morphology, which significantly influence their dynamics and contributions to IAC. Those variations necessitate a two-group classification: Group 1 (G1) for spherical and distorted bubbles, and Group 2 (G2) for cap, slug, and churn-turbulent bubbles. This study evaluated the predictive performance of several existing two-group IAC models, including IAC models implemented in RELAP5 and TRACE codes, as well as Song and Hibiki's IAC model, using compiled experimental data from vertically downward two-phase flows. The results showed that none of these models could predict the IAC with sufficient accuracy. To address this issue, the contribution of G1 bubbles to the total IAC was analyzed to investigate the flow behaviors for downward two-phase flows. Then, a two-group drift-flux correlation specifically developed for downward flows was introduced to predict the void fractions of G1 and G2 bubbles. Then, a new two-group Sauter mean diameter (SMD) model was formulated by considering the interfacial geometry effect and flow orientation. By integrating the drift-flux correlation and SMD model, an advanced two-group IAC model was proposed for downward two-phase flows. The proposed model demonstrated significantly improved predictive performance for downward dispersed flows, achieving mean absolute relative errors of 31.0 %, 26.7 %, and 27.9 % for G1, G2, and total IACs, respectively.
{"title":"Two-group interfacial area concentration model for downward dispersed two-phase flows","authors":"Shaoting Jia, Takashi Hibiki","doi":"10.1016/j.pnucene.2026.106271","DOIUrl":"10.1016/j.pnucene.2026.106271","url":null,"abstract":"<div><div>The two-fluid model is widely employed in thermal-hydraulic simulations of gas-liquid two-phase flows. Its predictive capability is strongly dependent on the accurate estimation of the interfacial area concentration (IAC), a key parameter governing the transfer of mass, momentum, and energy across the interphase. In dispersed flows, bubbles vary widely in size and morphology, which significantly influence their dynamics and contributions to IAC. Those variations necessitate a two-group classification: Group 1 (G1) for spherical and distorted bubbles, and Group 2 (G2) for cap, slug, and churn-turbulent bubbles. This study evaluated the predictive performance of several existing two-group IAC models, including IAC models implemented in RELAP5 and TRACE codes, as well as Song and Hibiki's IAC model, using compiled experimental data from vertically downward two-phase flows. The results showed that none of these models could predict the IAC with sufficient accuracy. To address this issue, the contribution of G1 bubbles to the total IAC was analyzed to investigate the flow behaviors for downward two-phase flows. Then, a two-group drift-flux correlation specifically developed for downward flows was introduced to predict the void fractions of G1 and G2 bubbles. Then, a new two-group Sauter mean diameter (SMD) model was formulated by considering the interfacial geometry effect and flow orientation. By integrating the drift-flux correlation and SMD model, an advanced two-group IAC model was proposed for downward two-phase flows. The proposed model demonstrated significantly improved predictive performance for downward dispersed flows, achieving mean absolute relative errors of 31.0 %, 26.7 %, and 27.9 % for G1, G2, and total IACs, respectively.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106271"},"PeriodicalIF":3.2,"publicationDate":"2026-02-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189703","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 : 2026-02-01DOI: 10.1016/j.pnucene.2026.106261
M.A. Hernández-Ceballos , M. Sangiorgi , N. Conte , J.P. Bolívar
Understanding the atmospheric dispersion and transport of radioactive materials is crucial for assessing radiological exposure and potential health risks, as well as for optimizing radiological environmental impact assessment and radiological monitoring networks. The dispersion of radionuclides following an accidental release from a nuclear facility is highly influenced by local wind circulation patterns, yet these effects are often overlooked in routine atmospheric dispersion assessments. This study evaluates the role of simple wind circulation indices—stagnation, recirculation, and ventilation—in shaping the dispersion of radioactive material, demonstrating their relevance for nuclear safety planning. The analysis focuses on the Almaraz Nuclear Power Plant (ANPP), where 1.256 atmospheric dispersion simulations were conducted using the RIMPUFF model over a four-year period (2012–2015) under different meteorological conditions. Considering the existing set of 84 monitoring stations included in the EURDEP system in an area of 200 km around the ANPP, the influence of each local atmospheric process is analyzed and characterized by taking the TGDR maximum values reached, and the number of monitoring stations affected in each simulation. On average, results demonstrate that high stagnation confines radionuclide plumes near the source, with maximum TGDR reaching 0.005 μSv/h and affecting up to 14 monitoring stations. In contrast, high recirculation enhances local accumulation, leading to, on average, peaks of 0.035 μSv/h and reducing the number of stations impacted (12 monitoring stations). High ventilation conditions promote wider dispersion, with maximum TGDR of 0.002 μSv/h affecting 10 monitoring stations. Extreme cases of each atmospheric process are also analyzed, showing distinct effects on the spatial distribution of affected monitoring stations. These findings highlight that wind circulation indices, derived from routine meteorological data, offer a straightforward yet effective means of anticipating dispersion behaviour in emergency scenarios.
{"title":"Evaluating local wind circulation metrics for radionuclide transport and dispersion: A practical approach for radiological safety","authors":"M.A. Hernández-Ceballos , M. Sangiorgi , N. Conte , J.P. Bolívar","doi":"10.1016/j.pnucene.2026.106261","DOIUrl":"10.1016/j.pnucene.2026.106261","url":null,"abstract":"<div><div>Understanding the atmospheric dispersion and transport of radioactive materials is crucial for assessing radiological exposure and potential health risks, as well as for optimizing radiological environmental impact assessment and radiological monitoring networks. The dispersion of radionuclides following an accidental release from a nuclear facility is highly influenced by local wind circulation patterns, yet these effects are often overlooked in routine atmospheric dispersion assessments. This study evaluates the role of simple wind circulation indices—stagnation, recirculation, and ventilation—in shaping the dispersion of radioactive material, demonstrating their relevance for nuclear safety planning. The analysis focuses on the Almaraz Nuclear Power Plant (ANPP), where 1.256 atmospheric dispersion simulations were conducted using the RIMPUFF model over a four-year period (2012–2015) under different meteorological conditions. Considering the existing set of 84 monitoring stations included in the EURDEP system in an area of 200 km around the ANPP, the influence of each local atmospheric process is analyzed and characterized by taking the TGDR maximum values reached, and the number of monitoring stations affected in each simulation. On average, results demonstrate that high stagnation confines radionuclide plumes near the source, with maximum TGDR reaching 0.005 μSv/h and affecting up to 14 monitoring stations. In contrast, high recirculation enhances local accumulation, leading to, on average, peaks of 0.035 μSv/h and reducing the number of stations impacted (12 monitoring stations). High ventilation conditions promote wider dispersion, with maximum TGDR of 0.002 μSv/h affecting 10 monitoring stations. Extreme cases of each atmospheric process are also analyzed, showing distinct effects on the spatial distribution of affected monitoring stations. These findings highlight that wind circulation indices, derived from routine meteorological data, offer a straightforward yet effective means of anticipating dispersion behaviour in emergency scenarios.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106261"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189702","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 : 2026-01-28DOI: 10.1016/j.pnucene.2026.106276
Mou Wang , Gen Jiang , Kai Wang , Songbai Cheng , Wenzhong Zhou
The long-term performance evolution and fission gas release (FGR) behavior of liquid metal-cooled fast reactor (LMFR) fuel elements are crucial for reactor safety and radioactive source term assessment. In this paper, to address the deficiencies of fuel performance analysis models for LMFR, a multi-physics field-coupled fuel performance analysis program is developed by using the FAST and CAMPUS program architectures and integrating the key physics models of FEAST, KMC-fuel, and other fast reactor programs. The program contains core modules for thermal-physical analysis, FGR, and chemical element migration (oxygen/plutonium), and considers a joint oxide gain (JOG) formation module. The program adopts two-dimensional axisymmetric geometry modeling to enhance the computational efficiency. Based on the verification of the program simulation results in comparison with the irradiated data of the experimental reactor, the present model shows high accuracy in the prediction of fuel temperature field distribution, gap closure kinetics, and fission gas release share (average relative error with experimental data is significantly lower than that of the FEAST model), and that the overall program is able to simulate the evolution of the overall performance of the fuel element well. Based on the validation of the overall performance of the program, the study also analyzes the key role of line power and JOG formation on fuel performance. It is shown that the pellet thermodynamic temperature rises significantly with increasing fuel operating power, which exacerbates the FGR behavior and induces an increase in cladding stress. More critically, the formation of JOG enhances the thermal conductivity of the fuel gap, which changes the temperature field distribution of the pellet, the mechanical deformation of the material, etc., and is thus a key factor that should not be ignored in the process of accurately predicting the change in fuel performance.
{"title":"Modeling fuel behavior in liquid metal fast reactors: A multiphysics approach with JOG formation analysis","authors":"Mou Wang , Gen Jiang , Kai Wang , Songbai Cheng , Wenzhong Zhou","doi":"10.1016/j.pnucene.2026.106276","DOIUrl":"10.1016/j.pnucene.2026.106276","url":null,"abstract":"<div><div>The long-term performance evolution and fission gas release (FGR) behavior of liquid metal-cooled fast reactor (LMFR) fuel elements are crucial for reactor safety and radioactive source term assessment. In this paper, to address the deficiencies of fuel performance analysis models for LMFR, a multi-physics field-coupled fuel performance analysis program is developed by using the FAST and CAMPUS program architectures and integrating the key physics models of FEAST, KMC-fuel, and other fast reactor programs. The program contains core modules for thermal-physical analysis, FGR, and chemical element migration (oxygen/plutonium), and considers a joint oxide gain (JOG) formation module. The program adopts two-dimensional axisymmetric geometry modeling to enhance the computational efficiency. Based on the verification of the program simulation results in comparison with the irradiated data of the experimental reactor, the present model shows high accuracy in the prediction of fuel temperature field distribution, gap closure kinetics, and fission gas release share (average relative error with experimental data is significantly lower than that of the FEAST model), and that the overall program is able to simulate the evolution of the overall performance of the fuel element well. Based on the validation of the overall performance of the program, the study also analyzes the key role of line power and JOG formation on fuel performance. It is shown that the pellet thermodynamic temperature rises significantly with increasing fuel operating power, which exacerbates the FGR behavior and induces an increase in cladding stress. More critically, the formation of JOG enhances the thermal conductivity of the fuel gap, which changes the temperature field distribution of the pellet, the mechanical deformation of the material, etc., and is thus a key factor that should not be ignored in the process of accurately predicting the change in fuel performance.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106276"},"PeriodicalIF":3.2,"publicationDate":"2026-01-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146079638","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 : 2026-01-24DOI: 10.1016/j.pnucene.2026.106273
Cun Liu , Liangming Pan , Longxiang Zhu , Jiewen Deng
Bubble condensation, as a form of direct contact condensation, is widely employed in engineering applications due to its high heat transfer efficiency. However, current understanding of the condensation behavior of bubbles containing non-condensable gases remains limited, particularly regarding the mechanism by which non-condensable gases impede bubble heat and mass transfer. In this study, the Volume of Fluid (VOF) method coupled with the Lee model and a species transport model is used to numerically simulate the condensation process of vapor bubbles containing non-condensable gas. By systematically varying key parameters—including initial bubble volume, liquid subcooling, and vapor mass fraction—the influence of non-condensable gas on bubble heat and mass transfer behavior is investigated. Results indicate significant differences in condensation rate and morphological evolution between pure vapor bubbles and those containing non-condensable gas. During the condensation of bubbles with non-condensable gas, the initial condensation rate shows a weak linear correlation with vapor mass fraction, whereas a strong linear relationship emerges in later stages. The content of non-condensable gas notably affects bubble shape evolution and rise velocity. Specifically, condensation initially concentrates at the lateral surface of the bubble, then gradually shifts toward the top and bottom in later stages. Moreover, under conditions of higher subcooling and higher vapor mass fraction, the internal flow within the bubble exhibits distinct disordered characteristics.
气泡冷凝作为直接接触冷凝的一种形式，由于其具有较高的传热效率，在工程上得到了广泛的应用。然而，目前对含有不可冷凝气体的气泡的冷凝行为的理解仍然有限，特别是关于不可冷凝气体阻碍气泡传热和传质的机制。本文采用流体体积法（Volume of Fluid， VOF）与Lee模型和物种输运模型相结合的方法，对含不可凝气体的蒸汽泡的凝结过程进行了数值模拟。通过系统地改变初始气泡体积、液体过冷度和蒸汽质量分数等关键参数，研究了不凝性气体对气泡传热传质行为的影响。结果表明，纯汽泡和含不凝气体的汽泡在凝结速率和形态演化上存在显著差异。在气泡与非可凝气体的冷凝过程中，初始冷凝速率与水蒸气质量分数呈弱线性相关，而后期则呈现强线性关系。不凝性气体含量对气泡形状演化和上升速度有显著影响。具体来说，凝结最初集中在气泡的侧面，然后在后期逐渐向顶部和底部转移。在过冷度和蒸气质量分数较高的条件下，气泡内部流动表现出明显的无序特征。
{"title":"Study on the interface characteristics of bubble condensation with non-condensable gas","authors":"Cun Liu , Liangming Pan , Longxiang Zhu , Jiewen Deng","doi":"10.1016/j.pnucene.2026.106273","DOIUrl":"10.1016/j.pnucene.2026.106273","url":null,"abstract":"<div><div>Bubble condensation, as a form of direct contact condensation, is widely employed in engineering applications due to its high heat transfer efficiency. However, current understanding of the condensation behavior of bubbles containing non-condensable gases remains limited, particularly regarding the mechanism by which non-condensable gases impede bubble heat and mass transfer. In this study, the Volume of Fluid (VOF) method coupled with the Lee model and a species transport model is used to numerically simulate the condensation process of vapor bubbles containing non-condensable gas. By systematically varying key parameters—including initial bubble volume, liquid subcooling, and vapor mass fraction—the influence of non-condensable gas on bubble heat and mass transfer behavior is investigated. Results indicate significant differences in condensation rate and morphological evolution between pure vapor bubbles and those containing non-condensable gas. During the condensation of bubbles with non-condensable gas, the initial condensation rate shows a weak linear correlation with vapor mass fraction, whereas a strong linear relationship emerges in later stages. The content of non-condensable gas notably affects bubble shape evolution and rise velocity. Specifically, condensation initially concentrates at the lateral surface of the bubble, then gradually shifts toward the top and bottom in later stages. Moreover, under conditions of higher subcooling and higher vapor mass fraction, the internal flow within the bubble exhibits distinct disordered characteristics.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106273"},"PeriodicalIF":3.2,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146079637","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 : 2026-01-24DOI: 10.1016/j.pnucene.2026.106274
Qingyin Zeng , Cheng Peng , Jiang Wu , Jian Deng
While existing experimental investigations and numerical simulations have preliminarily elucidated the multi-factor coupled mechanism governing Condensation-Induced-Water-Hammer (CIWH), the conventional threshold-based alarm approach predominantly relies on single-parameter criteria, e.g. typically pressure, which fails to effectively capture the early warning signs of CIWH. Concurrently, the genetic algorithm, recognized for its applicability in complex systems, remains relatively underutilized in the rapid identification of hazardous operating conditions, resulting in a technical gap where mechanistic understanding is disconnected from practical prevention and control requirements. This study investigates CIWH in the feedwater pipelines of the secondary circuit deaerator in nuclear power plants, where subcooled feedwater mixes with steam. A numerical simulation method based on the NUMAP code is developed to capture the underlying dynamics. Pilot study reveals that feedwater flow rate and pipe diameter jointly regulate the balance between inertial and frictional forces. This interaction shapes flow distribution and void fraction, which in turn influence the intensity of CIWH. Building upon this insight, a systematic analysis is conducted to quantify the effect of feedwater flow rate and pipe diameter on pressure variation rates, with peak values occurring at approximately 84.5 kg/s and 506 mm, respectively. To further identify hazardous operating conditions, a genetic algorithm (GA) is employed with different fitness functions. The results demonstrate that the relative change rate outperforms other metrics, whereas the mean absolute change and standard deviation show certain deviations, and the coefficient of variation is the least effective. This study confirms the effectiveness of the genetic algorithm in identifying hazardous operating conditions of CIWH under complex coupled scenarios and provides a feasible approach for predictive safety control and operational risk assessment in nuclear power plants.
{"title":"Diagnosis of water hammer conditions in multi-pipeline system using data-driven and genetic algorithm approach","authors":"Qingyin Zeng , Cheng Peng , Jiang Wu , Jian Deng","doi":"10.1016/j.pnucene.2026.106274","DOIUrl":"10.1016/j.pnucene.2026.106274","url":null,"abstract":"<div><div>While existing experimental investigations and numerical simulations have preliminarily elucidated the multi-factor coupled mechanism governing Condensation-Induced-Water-Hammer (CIWH), the conventional threshold-based alarm approach predominantly relies on single-parameter criteria, e.g. typically pressure, which fails to effectively capture the early warning signs of CIWH. Concurrently, the genetic algorithm, recognized for its applicability in complex systems, remains relatively underutilized in the rapid identification of hazardous operating conditions, resulting in a technical gap where mechanistic understanding is disconnected from practical prevention and control requirements. This study investigates CIWH in the feedwater pipelines of the secondary circuit deaerator in nuclear power plants, where subcooled feedwater mixes with steam. A numerical simulation method based on the NUMAP code is developed to capture the underlying dynamics. Pilot study reveals that feedwater flow rate and pipe diameter jointly regulate the balance between inertial and frictional forces. This interaction shapes flow distribution and void fraction, which in turn influence the intensity of CIWH. Building upon this insight, a systematic analysis is conducted to quantify the effect of feedwater flow rate and pipe diameter on pressure variation rates, with peak values occurring at approximately 84.5 kg/s and 506 mm, respectively. To further identify hazardous operating conditions, a genetic algorithm (GA) is employed with different fitness functions. The results demonstrate that the relative change rate outperforms other metrics, whereas the mean absolute change and standard deviation show certain deviations, and the coefficient of variation is the least effective. This study confirms the effectiveness of the genetic algorithm in identifying hazardous operating conditions of CIWH under complex coupled scenarios and provides a feasible approach for predictive safety control and operational risk assessment in nuclear power plants.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106274"},"PeriodicalIF":3.2,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039079","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 : 2026-01-24DOI: 10.1016/j.pnucene.2026.106263
Meiqi Song , Zuokai Chen , Jianhua Xia , Haozhe Li , Wei Xu , Xiaojing Liu
In nuclear power system, encountering the post-dryout heat transfer region can lead to severe heat transfer deterioration. Therefore, it is of great importance to give accurate prediction to post-dryout heat transfer. This study developed a new Physics-Embedded Machine Learning (PEML) framework to predict post-dryout heat transfer, addressing the limitations of traditional "black box" models by integrating physical constraints. Thirteen independent dimensionless parameters (e.g., , ), i.e., input features, and Nusselt number are derived to present the physical heat transfer mechanism. The is embedded into the loss function in proportionality or subtractive relationship, i.e., PEML(Nuexp/Nuc) and PEML(Nuexp-Nuc). The prediction capability of PEML models are better than traditional correlations. The PEML(Nuexp/Nuc) model achieves the best prediction capability with the mean error of 0.0005 and RMS error of 0.007 on the testing dataset from Becker's PDO experiments. It is indicated that increasing the number of input features generally improved model performance, especially the generalizability. The PEML framework successfully embeds heat transfer physics, bridging data-driven models and physical insights, offering a robust prediction tool for heat transfer.
{"title":"Prediction of post-dryout heat transfer based on physics-embedded machine learning with Bayesian optimization algorithm","authors":"Meiqi Song , Zuokai Chen , Jianhua Xia , Haozhe Li , Wei Xu , Xiaojing Liu","doi":"10.1016/j.pnucene.2026.106263","DOIUrl":"10.1016/j.pnucene.2026.106263","url":null,"abstract":"<div><div>In nuclear power system, encountering the post-dryout heat transfer region can lead to severe heat transfer deterioration. Therefore, it is of great importance to give accurate prediction to post-dryout heat transfer. This study developed a new Physics-Embedded Machine Learning (PEML) framework to predict post-dryout heat transfer, addressing the limitations of traditional \"black box\" models by integrating physical constraints. Thirteen independent dimensionless parameters (e.g., <span><math><mrow><mi>R</mi><msub><mi>e</mi><mtext>TP</mtext></msub></mrow></math></span>, <span><math><mrow><msub><mi>Pr</mi><mi>w</mi></msub></mrow></math></span>), i.e., input features, and Nusselt number <span><math><mrow><mi>N</mi><msub><mi>u</mi><mi>c</mi></msub></mrow></math></span> are derived to present the physical heat transfer mechanism. The <span><math><mrow><mi>N</mi><msub><mi>u</mi><mi>c</mi></msub></mrow></math></span> is embedded into the loss function in proportionality or subtractive relationship, i.e., PEML(<em>Nu</em><sub>exp</sub>/<em>Nu</em><sub>c</sub>) and PEML(<em>Nu</em><sub>exp</sub>-<em>Nu</em><sub>c</sub>). The prediction capability of PEML models are better than traditional correlations. The PEML(<em>Nu</em><sub>exp</sub>/<em>Nu</em><sub>c</sub>) model achieves the best prediction capability with the mean error of 0.0005 and RMS error of 0.007 on the testing dataset from Becker's PDO experiments. It is indicated that increasing the number of input features generally improved model performance, especially the generalizability. The PEML framework successfully embeds heat transfer physics, bridging data-driven models and physical insights, offering a robust prediction tool for heat transfer.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106263"},"PeriodicalIF":3.2,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039078","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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