Pub Date : 2025-02-22DOI: 10.1016/j.anucene.2025.111260
Inyup Kim, Taesuk Oh, Yonghee Kim
This paper presents the development, optimization, and verification of a depletion module integrated into the iMC Monte Carlo code. Several techniques are implemented to improve the performance and accuracy of the iMC depletion module. In addition, the nuclide control for the depletion of the molten salt reactors is developed. The performance of the depletion module is rigorously assessed through comprehensive code-to-code comparisons with the pre-validated Monte Carlo code Serpent. The evaluation encompasses three distinct depletion scenarios: a single PWR fuel pin, a single SFR fuel pin, and VERA benchmarks. Furthermore, the analysis extends to a simplified molten salt reactor experiment (MSRE) model, incorporating nuclide removal techniques. Comparisons focus on burnup-dependent infinite multiplication factors (kinf) and nuclide densities of actinides and fission products. Results demonstrate both the high accuracy and enhanced efficiency of the iMC Monte Carlo code’s depletion module, marking a significant advancement in advanced reactor analysis capabilities.
{"title":"Development and verification of depletion capabilities in the iMC Monte Carlo code","authors":"Inyup Kim, Taesuk Oh, Yonghee Kim","doi":"10.1016/j.anucene.2025.111260","DOIUrl":"10.1016/j.anucene.2025.111260","url":null,"abstract":"<div><div>This paper presents the development, optimization, and verification of a depletion module integrated into the iMC Monte Carlo code. Several techniques are implemented to improve the performance and accuracy of the iMC depletion module. In addition, the nuclide control for the depletion of the molten salt reactors is developed. The performance of the depletion module is rigorously assessed through comprehensive code-to-code comparisons with the pre-validated Monte Carlo code Serpent. The evaluation encompasses three distinct depletion scenarios: a single PWR fuel pin, a single SFR fuel pin, and VERA benchmarks. Furthermore, the analysis extends to a simplified molten salt reactor experiment (MSRE) model, incorporating nuclide removal techniques. Comparisons focus on burnup-dependent infinite multiplication factors (<em>k<sub>inf</sub></em>) and nuclide densities of actinides and fission products. Results demonstrate both the high accuracy and enhanced efficiency of the iMC Monte Carlo code’s depletion module, marking a significant advancement in advanced reactor analysis capabilities.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111260"},"PeriodicalIF":1.9,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143464155","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111283
Shilei Han , Pengfei Liu , Gang Wang, Bo Kuang, Yanhua Yang
The In-Vessel Retention (IVR) strategy has been widely applied in the existing and newly design Light Water Reactor (LWR). To give a sufficient safety margin for the further design of the large-scale reactors, the enhancement of the Critical Heat Flux (CHF) should be further studied. Porous coating technology is known as an effective CHF enhancement method. In this paper, the REPEC-III facility is modified to adapt the CHF enhancement conditions with higher heating power and wall temperature. The REPEC-III facility has 1:1 height ratio with the prototypic External Reactor Vessel Cooling (ERVC) environment. The flow channel is designed as a curved rectangular channel and the area ratio is 1:100 to the prototypic ERVC flow channel. The applicability of the porous coatings under the IVR-ERVC conditions are evaluated in this study. The cold spray technology is applied to prepare the porous coatings and prevent surface damage. The porous layer is composed of the dense basal layer and porous layer. The comparisons of the boiling phenomena are analyzed. On the porous surface, the larger amplitude and lower frequency of the pressure difference oscillation are observed and mean the more vapor slugs and the longer vapor period. During the experiments, the temperatures of the heating block with the porous surface are overall higher than the temperatures with the fresh surface. The heat transfer capability is worsened by the thermal resistance and improved liquid replenishment, which leads to a higher wall superheat under the same heat flux. The CHF is enhanced by the porous coatings and the enhancement effect is related to the orientations. The maximum percentage of the CHF enhancement is 43% at 87°. The intense turbulent induced by the more vapor slugs and the capillary wicking in the porous layer are beneficial to the liquid replenishment and enhance the CHF. Through the full-height experimental analysis, the porous coating technology is an effective method to improve the safety margin of the IVR strategy. The application of the porous coatings on the Reactor Pressure Vessel (RPV) outer wall are still needed to be further studied.
{"title":"Nucleate boiling heat transfer and CHF enhancement with porous surface coatings on the RPV outer wall","authors":"Shilei Han , Pengfei Liu , Gang Wang, Bo Kuang, Yanhua Yang","doi":"10.1016/j.anucene.2025.111283","DOIUrl":"10.1016/j.anucene.2025.111283","url":null,"abstract":"<div><div>The In-Vessel Retention (IVR) strategy has been widely applied in the existing and newly design Light Water Reactor (LWR). To give a sufficient safety margin for the further design of the large-scale reactors, the enhancement of the Critical Heat Flux (CHF) should be further studied. Porous coating technology is known as an effective CHF enhancement method. In this paper, the REPEC-III facility is modified to adapt the CHF enhancement conditions with higher heating power and wall temperature. The REPEC-III facility has 1:1 height ratio with the prototypic External Reactor Vessel Cooling (ERVC) environment. The flow channel is designed as a curved rectangular channel and the area ratio is 1:100 to the prototypic ERVC flow channel. The applicability of the porous coatings under the IVR-ERVC conditions are evaluated in this study. The cold spray technology is applied to prepare the porous coatings and prevent surface damage. The porous layer is composed of the dense basal layer and porous layer. The comparisons of the boiling phenomena are analyzed. On the porous surface, the larger amplitude and lower frequency of the pressure difference oscillation are observed and mean the more vapor slugs and the longer vapor period. During the experiments, the temperatures of the heating block with the porous surface are overall higher than the temperatures with the fresh surface. The heat transfer capability is worsened by the thermal resistance and improved liquid replenishment, which leads to a higher wall superheat under the same heat flux. The CHF is enhanced by the porous coatings and the enhancement effect is related to the orientations. The maximum percentage of the CHF enhancement is 43% at 87°. The intense turbulent induced by the more vapor slugs and the capillary wicking in the porous layer are beneficial to the liquid replenishment and enhance the CHF. Through the full-height experimental analysis, the porous coating technology is an effective method to improve the safety margin of the IVR strategy. The application of the porous coatings on the Reactor Pressure Vessel (RPV) outer wall are still needed to be further studied.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111283"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143454360","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111249
Laurent Laborde, Benoît Migot
Loss of cooling in a Spent Fuel Pool (SFP) of a nuclear power plant can lead to the melting of fuel assemblies and to strong radiological consequences to the environment. In order to study the first phases of such accidents, up to the fuel assemblies uncovery, the DENOPI project was launched by the French Institute for Radiation Protection and Nuclear Safety (IRSN1) supported and funded by the French Government and partners. Among the different facilities developed in the project, the MIDI facility aims at studying the complex thermal-hydraulics phenomena occurring in a large water pool heated from the bottom by electrical rods arranged in dedicated racks. MIDI is scaled by homothety to a typical French SFP. Different assembly arrangements (loading patterns) have been tested at different power levels, with either uniform power repartition, or hot and cold cells. In each test, the water level and temperatures at different elevations are followed, as well as mass flow rate entering each fuel rack. These experimental results also provide relevant data for the analysis and understanding of large natural convection loops that are expected in immersed passive heat removal systems of Small Modular Reactors. The forthcoming OECD/NEA POLCA project aims to extend such results database, in particular to assess the capability of thermal-hydraulics codes to reproduce the main tendencies of these experimental results.
The ASTEC code developed by IRSN is a system code dedicated to the simulation of major accidents in nuclear facilities that may lead to the release of radiological material. Recent works within the MUSA European project have shown the importance of reducing models uncertainties in the first phases of the accident, during the pool dewatering. In this paper, first simulations of MIDI tests are performed with ASTEC in order to assess and improve the capability of ASTEC to simulate the dewatering of a large water pool such as a SFP during a loss-of-cooling accident. Simulations are performed for a selection of MIDI tests with different heating patterns and power levels. Different models of subcooled boiling models from the literature are tested in ASTEC, stressing the key role of these models for an accurate prediction of the experimental flow.
{"title":"ASTEC validation of SFP dewatering using results from the DENOPI project","authors":"Laurent Laborde, Benoît Migot","doi":"10.1016/j.anucene.2025.111249","DOIUrl":"10.1016/j.anucene.2025.111249","url":null,"abstract":"<div><div>Loss of cooling in a Spent Fuel Pool (SFP) of a nuclear power plant can lead to the melting of fuel assemblies and to strong radiological consequences to the environment. In order to study the first phases of such accidents, up to the fuel assemblies uncovery, the DENOPI project was launched by the French Institute for Radiation Protection and Nuclear Safety (IRSN<span><span><sup>1</sup></span></span>) supported and funded by the French Government and partners. Among the different facilities developed in the project, the MIDI facility aims at studying the complex thermal-hydraulics phenomena occurring in a large water pool heated from the bottom by electrical rods arranged in dedicated racks. MIDI is scaled by homothety to a typical French SFP. Different assembly arrangements (loading patterns) have been tested at different power levels, with either uniform power repartition, or hot and cold cells. In each test, the water level and temperatures at different elevations are followed, as well as mass flow rate entering each fuel rack. These experimental results also provide relevant data for the analysis and understanding of large natural convection loops that are expected in immersed passive heat removal systems of Small Modular Reactors. The forthcoming OECD/NEA POLCA project aims to extend such results database, in particular to assess the capability of thermal-hydraulics codes to reproduce the main tendencies of these experimental results.</div><div>The ASTEC code developed by IRSN is a system code dedicated to the simulation of major accidents in nuclear facilities that may lead to the release of radiological material. Recent works within the MUSA European project have shown the importance of reducing models uncertainties in the first phases of the accident, during the pool dewatering. In this paper, first simulations of MIDI tests are performed with ASTEC in order to assess and improve the capability of ASTEC to simulate the dewatering of a large water pool such as a SFP during a loss-of-cooling accident. Simulations are performed for a selection of MIDI tests with different heating patterns and power levels. Different models of subcooled boiling models from the literature are tested in ASTEC, stressing the key role of these models for an accurate prediction of the experimental flow.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111249"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143464154","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111286
Long Yun, Xu Yuan, Guo Xi’an, Zhang Mingyu
In this paper, proper orthogonal decomposition (POD) technique is used to analyze the internal flow excitation characteristics of the reactor coolant pump. The stable operation of the reactor coolant pump (RCP), a critical component of nuclear power plants, is essential for maintaining reactor core cooling. The influence of the lower chamber of the steam generator on the pump inlet conditions is considered. Through numerical simulation and feature extraction techniques, the flow patterns and dynamic behaviors of key components such as impeller, diffuser and casing are analyzed in depth, and the multi-transient data of RCP are successfully processed. The POD analysis identifies the dominant energy structures within the flow field, offering insights into the primary flow characteristics. Studies have shown that POD technology can not only identify and explain complex flow phenomena under non-uniform inflow conditions, but also significantly improve the performance improvement and fault prevention capabilities of reactor coolant pumps.
{"title":"Analysis of internal flow excitation characteristics of reactor coolant pump based on POD","authors":"Long Yun, Xu Yuan, Guo Xi’an, Zhang Mingyu","doi":"10.1016/j.anucene.2025.111286","DOIUrl":"10.1016/j.anucene.2025.111286","url":null,"abstract":"<div><div>In this paper, proper orthogonal decomposition (POD) technique is used to analyze the internal flow excitation characteristics of the reactor coolant pump. The stable operation of the reactor coolant pump (RCP), a critical component of nuclear power plants, is essential for maintaining reactor core cooling. The influence of the lower chamber of the steam generator on the pump inlet conditions is considered. Through numerical simulation and feature extraction techniques, the flow patterns and dynamic behaviors of key components such as impeller, diffuser and casing are analyzed in depth, and the multi-transient data of RCP are successfully processed. The POD analysis identifies the dominant energy structures within the flow field, offering insights into the primary flow characteristics. Studies have shown that POD technology can not only identify and explain complex flow phenomena under non-uniform inflow conditions, but also significantly improve the performance improvement and fault prevention capabilities of reactor coolant pumps.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111286"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143454874","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111261
Yue Song , Meng Lv , Heng Xie
Curved U-Si/Al dispersion fuel is a promising candidate for the conversion of high-performance research reactors, particularly for high burnup conditions. In this study, a three-dimensional finite element model was developed to investigate the thermo-mechanical behavior of U-Si/Al fuel plates under irradiation, with a focus on the effects of curvature and creep. The model incorporates thermal expansion, irradiation-induced swelling, and creep, to simulate the performance of fuel plates under varying curvatures. The results demonstrate that: (1) the temperature distribution is very consistent in different curvatures regardless of creep; (2) creep significantly relaxes stress in the cladding, reducing the maximum Mises stress, while in the absence of creep, curvature has a greater impact on stress distribution; (3) the radial displacement of fuel plates is significantly influenced by creep, especially in plates with larger curvature radii. These findings are crucial for ensuring irradiation safety and optimizing the design of U-Si/Al fuel elements.
{"title":"Effects of the curvatures and creep on the thermo-mechanical behaviors in curved U-Si/Al dispersion fuel plates under irradiation condition","authors":"Yue Song , Meng Lv , Heng Xie","doi":"10.1016/j.anucene.2025.111261","DOIUrl":"10.1016/j.anucene.2025.111261","url":null,"abstract":"<div><div>Curved U-Si/Al dispersion fuel is a promising candidate for the conversion of high-performance research reactors, particularly for high burnup conditions. In this study, a three-dimensional finite element model was developed to investigate the thermo-mechanical behavior of U-Si/Al fuel plates under irradiation, with a focus on the effects of curvature and creep. The model incorporates thermal expansion, irradiation-induced swelling, and creep, to simulate the performance of fuel plates under varying curvatures. The results demonstrate that: (1) the temperature distribution is very consistent in different curvatures regardless of creep; (2) creep significantly relaxes stress in the cladding, reducing the maximum Mises stress, while in the absence of creep, curvature has a greater impact on stress distribution; (3) the radial displacement of fuel plates is significantly influenced by creep, especially in plates with larger curvature radii. These findings are crucial for ensuring irradiation safety and optimizing the design of U-Si/Al fuel elements.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111261"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143454359","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111274
Alaa A. Mahmoud , Alaa A. El-Sayed , Ayman M. Aboraya , Islam N. Fathy , Mohamed A. Abouelnour , Bassam A. Tayeh , Islam M. Nabil
Global warming and resource depletion necessitate sustainable concrete production. This study examines the use of waste construction powders as partial cement replacements (up to 9 %) in concrete. The waste powders studied were granite (WGP), marble (WMP), granodiorite (WGDP), and ceramic (WCP). The effects of these waste powders on the mechanical, microstructural properties, and radiation shielding capabilities of ordinary concrete were investigated. The concrete mixture that demonstrated the greatest compressive strength, a 25.6% increase after 28 days, contained a 9% WGP replacement ratio. The optimal tensile strength after 28 days, showing a 19.7% improvement, was achieved with a 7% WCP replacement rate. The concrete specimens with the highest compressive strength (containing waste powders) demonstrated enhanced radiation shielding capabilities compared to the control mix. For WCP concrete mix, the CMix linear attenuation values were the highest due to its high density and greatest Ti/Fe content. Fast-neutron removal worked best with the 9 % WGP.
{"title":"Investigating the effects of granite, marble, granodiorite, and ceramic waste powders on the physical, mechanical, and radiation shielding performance of sustainable concrete","authors":"Alaa A. Mahmoud , Alaa A. El-Sayed , Ayman M. Aboraya , Islam N. Fathy , Mohamed A. Abouelnour , Bassam A. Tayeh , Islam M. Nabil","doi":"10.1016/j.anucene.2025.111274","DOIUrl":"10.1016/j.anucene.2025.111274","url":null,"abstract":"<div><div>Global warming and resource depletion necessitate sustainable concrete production. This study examines the use of waste construction powders as partial cement replacements (up to 9 %) in concrete. The waste powders studied were granite (WGP), marble (WMP), granodiorite (WGDP), and ceramic (WCP). The effects of these waste powders on the mechanical, microstructural properties, and radiation shielding capabilities of ordinary concrete were investigated. The concrete mixture that demonstrated the greatest compressive strength, a 25.6% increase after 28 days, contained a 9% WGP replacement ratio. The optimal tensile strength after 28 days, showing a 19.7% improvement, was achieved with a 7% WCP replacement rate. The concrete specimens with the highest compressive strength (containing waste powders) demonstrated enhanced radiation shielding capabilities compared to the control mix. For WCP concrete mix, the CMix linear attenuation values were the highest due to its high density and greatest Ti/Fe content. Fast-neutron removal worked best with the 9 % WGP.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111274"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143454873","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111293
Xiangyu Chi , Yaru Li , Bin Zhao , Naihua Wang , Pei Yu , Jiazhi Liu
The volume control tank (VCT) is a vital equipment in the chemical and volume control system (RCV). Its retention effect on the influent seriously affects the accuracy and the response speed of the reactor coolant system (RCS) boron concentration. This paper aims to study the influence of inlet configurations on mixing and fluid retention in VCT using a numerical method and provide guidance for the design of VCT. Seven feasible inlet configurations are proposed. Detailed concentration and velocity profiles are provided to analyze mixing and fluid retention mechanisms in VCT. The results show that fluid retention is closely related to mixing, and the weaker the mixing, the more intense the fluid retention. Config.-D-H, Config.-D-T, and Config.-D-P are advisable choices when minimizing the deviation between the actual and the target RCS concentrations is urgent. Config.-S-L is preferable when the rapid response of RCS boron concentration is crucial.
{"title":"Influence of inlet configurations on mixing and fluid retention in VCT","authors":"Xiangyu Chi , Yaru Li , Bin Zhao , Naihua Wang , Pei Yu , Jiazhi Liu","doi":"10.1016/j.anucene.2025.111293","DOIUrl":"10.1016/j.anucene.2025.111293","url":null,"abstract":"<div><div>The volume control tank (VCT) is a vital equipment in the chemical and volume control system (RCV). Its retention effect on the influent seriously affects the accuracy and the response speed of the reactor coolant system (RCS) boron concentration. This paper aims to study the influence of inlet configurations on mixing and fluid retention in VCT using a numerical method and provide guidance for the design of VCT. Seven feasible inlet configurations are proposed. Detailed concentration and velocity profiles are provided to analyze mixing and fluid retention mechanisms in VCT. The results show that fluid retention is closely related to mixing, and the weaker the mixing, the more intense the fluid retention. <strong>Config.-D-H</strong>, <strong>Config.-D-T</strong>, and <strong>Config.-D-P</strong> are advisable choices when minimizing the deviation between the actual and the target RCS concentrations is urgent. <strong>Config.-S-L</strong> is preferable when the rapid response of RCS boron concentration is crucial.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111293"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143454875","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 : 2025-02-21DOI: 10.1016/j.anucene.2025.111250
Leonardo R.C. Moraes , Ricardo C. Barros , Hermes Alves Filho , Richard Vasques
In this work, we present an extension of the nonclassical transport model, namely the generalized linear Boltzmann equation (GLBE), to eigenvalue criticality problems. The GLBE is a generalization of the linear Boltzmann equation that allows the modeling of particle transport in random statistically homogeneous systems in which the free-path distribution function is non-exponential. This type of problem is referred to as a nonclassical transport problem. The model’s ability to accurately replicate the expected value for the system’s effective multiplication factor and the profile of the neutron scalar flux for both classical and nonclassical transport problems is analyzed.
{"title":"A nonclassical model to eigenvalue neutron transport calculations","authors":"Leonardo R.C. Moraes , Ricardo C. Barros , Hermes Alves Filho , Richard Vasques","doi":"10.1016/j.anucene.2025.111250","DOIUrl":"10.1016/j.anucene.2025.111250","url":null,"abstract":"<div><div>In this work, we present an extension of the nonclassical transport model, namely the generalized linear Boltzmann equation (GLBE), to eigenvalue criticality problems. The GLBE is a generalization of the linear Boltzmann equation that allows the modeling of particle transport in random statistically homogeneous systems in which the free-path distribution function <span><math><mrow><mi>p</mi><mrow><mo>(</mo><mi>Ω</mi><mo>,</mo><mi>s</mi><mo>)</mo></mrow></mrow></math></span> is non-exponential. This type of problem is referred to as a nonclassical transport problem. The model’s ability to accurately replicate the expected value for the system’s effective multiplication factor and the profile of the neutron scalar flux for both classical and nonclassical transport problems is analyzed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111250"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143464215","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 presents the development of oxygen sensors based on four electrode materials lanthanum strontium cobalt ferrite (LSCF), lanthanum strontium manganese (LSM), platinum (Pt) and silver (Ag), and yttria-stabilized zirconia (YSZ) solid electrolyte. The responsiveness, accuracy, stability and repeatability of the various oxygen sensors were evaluated. The findings indicate that the LSCF, LSM and Ag electrode oxygen sensors exhibit optimal operational suitability, whereas the Pt electrode oxygen sensor is suitable for environments exceeding 650 K. The performance of 5YSZ solid electrolytes is more susceptible to low temperature effects than that of 8YSZ solid electrolytes. And the output electromotive force (EMF) as a function of temperature for the liquid LBE in a saturated oxygen environment was obtained, show as: , within 473 ∼ 823 K. Furthermore, the oxygen solubility of the liquid LBE was determined by combining the aforementioned thermodynamic data, show as: , (CO,S in wt.%).
{"title":"Electrochemical equilibrium of air-referenced oxygen sensors in liquid LBE and analysis for determination of oxygen solubility","authors":"Ruixian Liang, Huiping Zhu, Hao Wu, Zhangpeng Guo, Haicai Lyu, Zulong Hao, Yang liu, Fang Liu, Fenglei Niu","doi":"10.1016/j.anucene.2025.111272","DOIUrl":"10.1016/j.anucene.2025.111272","url":null,"abstract":"<div><div>This paper presents the development of oxygen sensors based on four electrode materials lanthanum strontium cobalt ferrite (LSCF), lanthanum strontium manganese (LSM), platinum (Pt) and silver (Ag), and yttria-stabilized zirconia (YSZ) solid electrolyte. The responsiveness, accuracy, stability and repeatability of the various oxygen sensors were evaluated. The findings indicate that the LSCF, LSM and Ag electrode oxygen sensors exhibit optimal operational suitability, whereas the Pt electrode oxygen sensor is suitable for environments exceeding 650 K. The performance of 5YSZ solid electrolytes is more susceptible to low temperature effects than that of 8YSZ solid electrolytes. And the output electromotive force (EMF) as a function of temperature for the liquid LBE in a saturated oxygen environment was obtained, show as: <span><math><mrow><mi>E</mi><mo>=</mo><mn>1.1307</mn><mo>-</mo><mn>5.9443</mn><mo>×</mo><msup><mn>10</mn><mrow><mo>-</mo><mn>4</mn></mrow></msup><mi>T</mi></mrow></math></span>, within 473 ∼ 823 K. Furthermore, the oxygen solubility of the liquid LBE was determined by combining the aforementioned thermodynamic data, show as: <span><math><mrow><mi>ln</mi><msub><mi>C</mi><mrow><mi>O</mi><mo>,</mo><mi>S</mi></mrow></msub><mo>=</mo><mn>7.09</mn><mo>-</mo><mn>10922</mn><mo>/</mo><mi>T</mi></mrow></math></span>, (<em>C<sub>O,S</sub></em> in wt.%).</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111272"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143464214","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 study applies a Monte Carlo-based sensitivity calculation model for self-powered neutron detectors (SPND). The model employs the neutron energy spectrum in pressurized water reactors and the beta decay energy spectrum of rhodium as input sources. Additionally, the sensitivity calculation model incorporates space charge effects, self-shielding factors, and the core temperature of the reactor. Compared with the results of other researchers, the model is reliable and can be used for further calculation. The consistency calculation work for sensitivity primarily investigates the influence of emitter thickness, insulator thickness, collector thickness, emitter length, emitter eccentricity, and isotopic composition of emitter material on sensitivity consistency. The results indicate that variations in insulator thickness have a significant impact on sensitivity, whereas changes in emitter eccentricity and trace element content in the rhodium wire have minimal effects on sensitivity. This research provides critical data support for the commercial production of SPND.
{"title":"Study of the influence of manufacturing defects on the sensitivity of rhodium self-powered neutron detectors","authors":"Zhiqi Guo , Yuanji Zhang , Wenhua Yang , Dianwei Zhou , Ren Xiao , Zhan Li , Dingjun Zhu , Jianxiong Shao","doi":"10.1016/j.anucene.2025.111285","DOIUrl":"10.1016/j.anucene.2025.111285","url":null,"abstract":"<div><div>This study applies a Monte Carlo-based sensitivity calculation model for self-powered neutron detectors (SPND). The model employs the neutron energy spectrum in pressurized water reactors and the beta decay energy spectrum of rhodium as input sources. Additionally, the sensitivity calculation model incorporates space charge effects, self-shielding factors, and the core temperature of the reactor. Compared with the results of other researchers, the model is reliable and can be used for further calculation. The consistency calculation work for sensitivity primarily investigates the influence of emitter thickness, insulator thickness, collector thickness, emitter length, emitter eccentricity, and isotopic composition of emitter material on sensitivity consistency. The results indicate that variations in insulator thickness have a significant impact on sensitivity, whereas changes in emitter eccentricity and trace element content in the rhodium wire have minimal effects on sensitivity. This research provides critical data support for the commercial production of SPND.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111285"},"PeriodicalIF":1.9,"publicationDate":"2025-02-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143454358","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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