Pub Date : 2025-03-04DOI: 10.1016/j.anucene.2025.111256
Tatsuhiko Ogawa
A novel robust method has been developed to simulate the performance of composite neutron sources composed of an alpha-emitting actinide and a light nucleus with low neutron separation energy. This method is based on the JENDL-5 cross-section data library and the Monte-Carlo radiation transport code PHITS. In contrast to previously devised methods, this approach can predict various quantities of the sources, such as actinide grain size dependence, absolute neutron emission intensity, energy spectra of neutrons and parasitic photons, neutron multiplicity, and time structure, with little approximation.
The accurate calculation of stopping power of alpha rays in actinide grains and light elements, as well as the use of (, n) reaction evaluated cross sections, which is one of the unique features of PHITS Ver.3.34 and its later versions, are the essences of the method. This method allows for the calculation of quantities important for practical applications, such as detection signal frequency, coincidence event rate, and the impact of parasitic gamma-rays.
{"title":"Prediction of composite neutron source spectra by combination of JENDL-5 and PHITS","authors":"Tatsuhiko Ogawa","doi":"10.1016/j.anucene.2025.111256","DOIUrl":"10.1016/j.anucene.2025.111256","url":null,"abstract":"<div><div>A novel robust method has been developed to simulate the performance of composite neutron sources composed of an alpha-emitting actinide and a light nucleus with low neutron separation energy. This method is based on the JENDL-5 cross-section data library and the Monte-Carlo radiation transport code PHITS. In contrast to previously devised methods, this approach can predict various quantities of the sources, such as actinide grain size dependence, absolute neutron emission intensity, energy spectra of neutrons and parasitic photons, neutron multiplicity, and time structure, with little approximation.</div><div>The accurate calculation of stopping power of alpha rays in actinide grains and light elements, as well as the use of (<span><math><mi>α</mi></math></span>, n) reaction evaluated cross sections, which is one of the unique features of PHITS Ver.3.34 and its later versions, are the essences of the method. This method allows for the calculation of quantities important for practical applications, such as detection signal frequency, coincidence event rate, and the impact of parasitic gamma-rays.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111256"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534218","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 : 2025-03-04DOI: 10.1016/j.anucene.2025.111322
Yao Yao , Tao Zhou , Jianyu Tang , Dongli Huang , Zefeng Wang
Natural circulation flow instability is an important phenomenon for nuclear system safety, especially in narrow channels which have a broad application prospect. However, the small gap of narrow channels can easily lead to the narrow-space effect, resulting in mechanisms and performance of instability different from those in conventional channels. Medium-to-high pressure is a common condition in nuclear systems. Therefore, it is necessary to study flow instability in narrow channels under medium-to-high pressure. The ultimate goal is to effectively avoid the instability and improve the safety of nuclear systems. This manuscript investigates the characteristics of natural circulation instability in a narrow channel with deionized water under medium-to-high pressure (7.0 MPa-15.0 MPa). Results provide new insights into the safety and reliability of nuclear reactor cooling systems, supporting the establishment of more accurate and reliable natural circulation systems. A numerical study establishes a model of natural circulation loop including a narrow channel via system code RELAP5 based on a well-validated natural circulation test facility. The instability development is divided into three stages, natural circulation stable flow, instability flow (no reverse flow), and instability flow (reverse flow) or periodic dryout instability, according to the amplitudes and periods of mass flow rate. The proposed RELAP5 model demonstrates the relationship between mass flow rate and pressure drop on a time scale of a few seconds in every stage. Mass flow rate, exit quality, two-phase length, and flow regime are examined throughout all stages. Results indicate that increasing system pressure and inlet subcooling have a significant inhibitory effect on instability oscillation.
{"title":"Numerical study on flow instability development of natural circulation in narrow channel under medium-to-high pressure","authors":"Yao Yao , Tao Zhou , Jianyu Tang , Dongli Huang , Zefeng Wang","doi":"10.1016/j.anucene.2025.111322","DOIUrl":"10.1016/j.anucene.2025.111322","url":null,"abstract":"<div><div>Natural circulation flow instability is an important phenomenon for nuclear system safety, especially in narrow channels which have a broad application prospect. However, the small gap of narrow channels can easily lead to the narrow-space effect, resulting in mechanisms and performance of instability different from those in conventional channels. Medium-to-high pressure is a common condition in nuclear systems. Therefore, it is necessary to study flow instability in narrow channels under medium-to-high pressure. The ultimate goal is to effectively avoid the instability and improve the safety of nuclear systems. This manuscript investigates the characteristics of natural circulation instability in a narrow channel with deionized water under medium-to-high pressure (7.0 MPa-15.0 MPa). Results provide new insights into the safety and reliability of nuclear reactor cooling systems, supporting the establishment of more accurate and reliable natural circulation systems. A numerical study establishes a model of natural circulation loop including a narrow channel via system code RELAP5 based on a well-validated natural circulation test facility. The instability development is divided into three stages, natural circulation stable flow, instability flow (no reverse flow), and instability flow (reverse flow) or periodic dryout instability, according to the amplitudes and periods of mass flow rate. The proposed RELAP5 model demonstrates the relationship between mass flow rate and pressure drop on a time scale of a few seconds in every stage. Mass flow rate, exit quality, two-phase length, and flow regime are examined throughout all stages. Results indicate that increasing system pressure and inlet subcooling have a significant inhibitory effect on instability oscillation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111322"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534219","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 : 2025-03-04DOI: 10.1016/j.anucene.2025.111320
Ying Cao , Weishi Wan , Chong Zhou
Liquid-fueled molten salt reactors (LFMSR) utilize internally heated fuel-salt compounds as the working fluid for the first time and the heat source and heat sink conditions are crucial for the natural circulation phenomena. This paper explored the steady-state behaviors of natural circulation of LFMSR and analyzed the impact of internal heat. By incorporating a heat source term into the liquid phase, the natural circulation governing differential equations were modified and solved using a self-developed Python code. Then, the homogeneously volumetric heat model of LFMSR was compared with the core central heat model of a traditional solid fuel reactor, the LFMSR exhibited a smaller steady-state natural circulation flow reducing the natural circulation thermal effect. Subsequently, an extended RELAP5/SCDAPSIM/MOD4.0 code for LFMSR was applied to verify the self-developed code, and their quantitative steady-state parameters showed a good match, which can be used for more complex system analyses in future works.
{"title":"Research on steady-state behavior of natural circulation with internally heated fluids of liquid-fueled molten salt reactor","authors":"Ying Cao , Weishi Wan , Chong Zhou","doi":"10.1016/j.anucene.2025.111320","DOIUrl":"10.1016/j.anucene.2025.111320","url":null,"abstract":"<div><div>Liquid-fueled molten salt reactors (LFMSR) utilize internally heated fuel-salt compounds as the working fluid for the first time and the heat source and heat sink conditions are crucial for the natural circulation phenomena. This paper explored the steady-state behaviors of natural circulation of LFMSR and analyzed the impact of internal heat. By incorporating a heat source term into the liquid phase, the natural circulation governing differential equations were modified and solved using a self-developed Python code. Then, the homogeneously volumetric heat model of LFMSR was compared with the core central heat model of a traditional solid fuel reactor, the LFMSR exhibited a smaller steady-state natural circulation flow reducing the natural circulation thermal effect. Subsequently, an extended RELAP5/SCDAPSIM/MOD4.0 code for LFMSR was applied to verify the self-developed code, and their quantitative steady-state parameters showed a good match, which can be used for more complex system analyses in future works.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111320"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534353","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 : 2025-03-04DOI: 10.1016/j.anucene.2025.111317
Qi Wang , Yu Zhong , Chenggang Yu , Jianhui Wu , Zhichao Wang , Wei Guo , Jingen Chen , Xiangzhou Cai
<div><div>As one of the prospective Generation IV reactor designs, molten salt reactors (MSRs) are receiving increasing attention due to its unique advantages. Graphite is a preferred moderator in an MSR with the excellent moderation capability, good compatibility with the fuel salt and the stability under neutron radiation. However, the narrow gaps filled with fuel salt are existed between adjacent graphite assemblies due to the fabrication tolerances, the irradiation-induced deformation, and the different thermal expansion coefficients between the support structure and graphite. As the fuel salt acts as both a heat source and coolant, the effect of narrow gaps on the component performance and their impact on graphite lifespan is necessary to be evaluated. This study analyzes the specific effects of narrow gaps between the adjacent graphite components in a round channel assembly (RCA) and round groove assembly (RGA) of a small modular molten salt reactor (SM-MSR), respectively. The results show that the performances of graphite assemblies with the narrow gaps significantly differ from those of ideal assemblies with no gaps. In terms of the component structure, the variations on the neutronic and thermohydraulic performances between RGA and RCA are primarily caused by the distinct positions of the narrow gap in each component. One the one hand, the heat flux distribution indicates that the component structures have a significant impact on the overall cooling capacity, where RGA exhibited a more than 19.98 % reduction in the heat removal through the primary channel to the graphite compared to RCA with a 0.5 mm narrow gap. The effective cooling range extends from inlet to the outlet with the increased narrow gap width. On the other hand, the flow exchange between the fuel salt in channels is enhanced in RGA due to the connectivity of the primary and narrow gap channels. Furthermore, the hotspot in the component is mitigated by the increase in the mass flow rate of fuel salt through the narrow gap flow channel, which is more prominent with the widening of the narrow gap. Therefore, the adverse effects on the temperature distribution caused by the narrow gap can be alleviated in RGA compared to RCA. Due to the enhanced flow when the narrow gap width is increased from 0.5 mm to 3 mm, the temperature gradient is decreased by 68.56 % and 62.38 % for RCA and RGA, respectively. The graphite lifetime is also evaluated under the considerations of narrow gap width and component structure. When the narrow gap width is expanded to be 3 mm, the graphite lifetime of RGA is observed to be 7.47 % longer than that of RCA. The component lifespan is slightly improved with the increased width of the narrow gap. For the narrow gap width of 0.5 mm, the lifespan of RCA and RGA is reduced by 30.88 % and 22.43 %, respectively, compared to the ideal assembly with no narrow gap. Furthermore, when the narrow gap width of RGA reached to be 3 mm, the graphite lifespan is superi
{"title":"Influence of narrow gaps between the adjacent components on graphite lifetime in molten salt reactors","authors":"Qi Wang , Yu Zhong , Chenggang Yu , Jianhui Wu , Zhichao Wang , Wei Guo , Jingen Chen , Xiangzhou Cai","doi":"10.1016/j.anucene.2025.111317","DOIUrl":"10.1016/j.anucene.2025.111317","url":null,"abstract":"<div><div>As one of the prospective Generation IV reactor designs, molten salt reactors (MSRs) are receiving increasing attention due to its unique advantages. Graphite is a preferred moderator in an MSR with the excellent moderation capability, good compatibility with the fuel salt and the stability under neutron radiation. However, the narrow gaps filled with fuel salt are existed between adjacent graphite assemblies due to the fabrication tolerances, the irradiation-induced deformation, and the different thermal expansion coefficients between the support structure and graphite. As the fuel salt acts as both a heat source and coolant, the effect of narrow gaps on the component performance and their impact on graphite lifespan is necessary to be evaluated. This study analyzes the specific effects of narrow gaps between the adjacent graphite components in a round channel assembly (RCA) and round groove assembly (RGA) of a small modular molten salt reactor (SM-MSR), respectively. The results show that the performances of graphite assemblies with the narrow gaps significantly differ from those of ideal assemblies with no gaps. In terms of the component structure, the variations on the neutronic and thermohydraulic performances between RGA and RCA are primarily caused by the distinct positions of the narrow gap in each component. One the one hand, the heat flux distribution indicates that the component structures have a significant impact on the overall cooling capacity, where RGA exhibited a more than 19.98 % reduction in the heat removal through the primary channel to the graphite compared to RCA with a 0.5 mm narrow gap. The effective cooling range extends from inlet to the outlet with the increased narrow gap width. On the other hand, the flow exchange between the fuel salt in channels is enhanced in RGA due to the connectivity of the primary and narrow gap channels. Furthermore, the hotspot in the component is mitigated by the increase in the mass flow rate of fuel salt through the narrow gap flow channel, which is more prominent with the widening of the narrow gap. Therefore, the adverse effects on the temperature distribution caused by the narrow gap can be alleviated in RGA compared to RCA. Due to the enhanced flow when the narrow gap width is increased from 0.5 mm to 3 mm, the temperature gradient is decreased by 68.56 % and 62.38 % for RCA and RGA, respectively. The graphite lifetime is also evaluated under the considerations of narrow gap width and component structure. When the narrow gap width is expanded to be 3 mm, the graphite lifetime of RGA is observed to be 7.47 % longer than that of RCA. The component lifespan is slightly improved with the increased width of the narrow gap. For the narrow gap width of 0.5 mm, the lifespan of RCA and RGA is reduced by 30.88 % and 22.43 %, respectively, compared to the ideal assembly with no narrow gap. Furthermore, when the narrow gap width of RGA reached to be 3 mm, the graphite lifespan is superi","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111317"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143550104","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 : 2025-03-03DOI: 10.1016/j.anucene.2025.111294
Md Tanbirul Islam Rupam , Nahid Hasan , Md. Sheikh Rasel , Sumon Saha
The present study computationally investigates magnetohydrodynamic (MHD) mixed convective fluid circulation and entropy generation in a T-shaped open chamber containing a heat-generating and conducting cylinder. Ferrofluid is circulated through the enclosure by entering at the bottom and leaving from the top of both side openings. This study utilizes the finite element scheme to unravel the leading thermal energy and Navier-Stokes equations, employing suitable auxiliary conditions. This research aims to analyze the effects of governing non-dimensional governing and geometric parameters and explore the best thermo-fluid performance inside the enclosure. The geometrical and controlling parameters are the cylinder location in the vertical direction (δ = 0.6, 0.7, 0.8), Reynolds number (31.62 ≤ Re ≤ 316.23), Grashof number (103 ≤ Gr ≤ 105), Richardson number (0.1 ≤ Ri ≤ 10), Stuart number (0 ≤ N ≤ 3.16), Hartmann number (0 ≤ Ha ≤ 17.78), and Joule heating parameter (0 ≤ J ≤ 4.57 × 10−8). The outcomes of this investigation are assessed using numerical computations of the overall entropy generation within the enclosure, average Nusselt number along the edge of the heated cylinder, mean temperature of the solid cylinder, and thermal performance criterion for six distinct cases. Furthermore, a visual depiction of the fluid circulation and thermal fields is presented. Upon thorough examination, it becomes evident that elevated Reynolds and Grashof numbers result in increased heat transport and reduced entropy production. Moreover, the optimal vertical location of the cylinder is identified at 0.6 times the chamber height. The maximum Nusselt number is achieved in Case 1 (at fixed N and Gr), where a 26.78 % improvement can be obtained by adjusting the parameter values at δ = 0.6. The inclusive discoveries of the current study grasp the noteworthy potential for apprising the design of miscellaneous thermal systems, together with solar thermal collectors, nuclear reactor cooling, electronic cooling, etc.
{"title":"Magnetohydrodynamic conjugate mixed convection, Joule Heating, and entropy generation through a ferrofluid filled T-shaped open miniature chamber with a Heat-Generating circular rod","authors":"Md Tanbirul Islam Rupam , Nahid Hasan , Md. Sheikh Rasel , Sumon Saha","doi":"10.1016/j.anucene.2025.111294","DOIUrl":"10.1016/j.anucene.2025.111294","url":null,"abstract":"<div><div>The present study computationally investigates magnetohydrodynamic (MHD) mixed convective fluid circulation and entropy generation in a <em>T</em>-shaped open chamber containing a heat-generating and conducting cylinder. Ferrofluid is circulated through the enclosure by entering at the bottom and leaving from the top of both side openings. This study utilizes the finite element scheme to unravel the leading thermal energy and Navier-Stokes equations, employing suitable auxiliary conditions. This research aims to analyze the effects of governing non-dimensional governing and geometric parameters and explore the best thermo-fluid performance inside the enclosure. The geometrical and controlling parameters are the cylinder location in the vertical direction (<em>δ</em> = 0.6, 0.7, 0.8), Reynolds number (31.62 ≤ <em>Re</em> ≤ 316.23), Grashof number (10<sup>3</sup> ≤ <em>Gr</em> ≤ 10<sup>5</sup>), Richardson number (0.1 ≤ <em>Ri</em> ≤ 10), Stuart number (0 ≤ <em>N</em> ≤ 3.16), Hartmann number (0 ≤ <em>Ha</em> ≤ 17.78), and Joule heating parameter (0 ≤ <em>J</em> ≤ 4.57 × 10<sup>−8</sup>). The outcomes of this investigation are assessed using numerical computations of the overall entropy generation within the enclosure, average Nusselt number along the edge of the heated cylinder, mean temperature of the solid cylinder, and thermal performance criterion for six distinct cases. Furthermore, a visual depiction of the fluid circulation and thermal fields is presented. Upon thorough examination, it becomes evident that elevated Reynolds and Grashof numbers result in increased heat transport and reduced entropy production. Moreover, the optimal vertical location of the cylinder is identified at 0.6 times the chamber height. The maximum Nusselt number is achieved in Case 1 (at fixed <em>N</em> and <em>Gr</em>), where a 26.78 % improvement can be obtained by adjusting the parameter values at <em>δ</em> = 0.6. The inclusive discoveries of the current study grasp the noteworthy potential for apprising the design of miscellaneous thermal systems, together with solar thermal collectors, nuclear reactor cooling, electronic cooling, etc.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111294"},"PeriodicalIF":1.9,"publicationDate":"2025-03-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143529812","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 : 2025-03-01DOI: 10.1016/j.anucene.2025.111279
Gabriel C.G.R. da Silva , Carolina P. Naveira-Cotta , Kleber M. Lisboa , Renato M. Cotta , Jian Su
Despite the proven feasibility and cost-effectiveness of nuclear desalination in small modular reactors (SMRs), the exclusive use of waste heat for this purpose remains virtually unexplored. This work investigates coupling an SMR (NuScale) to a direct contact membrane distillation (DCMD) desalination plant with heat recovery and feed recycle. Both the reactor waste heat and the low-pressure (LP) steam extraction from the reactor turbine were considered as heat sources. The DCMD hollow fiber module was modeled as a porous medium with satisfactory accuracy. Key parameters affecting system performance, identified through factorial analysis, include membrane porosity, module length, feed superficial velocity, and fiber inner radius. Single and multiobjective optimization analyses revealed the feasibility of producing up to 3,810 m3/d of water without any reactor power loss, and up to 8,832 m3/d, with a 2.28 MWe power loss, using steam extraction. DCMD demonstrated competitiveness, especially when the reactor’s primary purpose is electricity generation.
{"title":"Assessment of nuclear desalination in a small modular reactor using membrane distillation","authors":"Gabriel C.G.R. da Silva , Carolina P. Naveira-Cotta , Kleber M. Lisboa , Renato M. Cotta , Jian Su","doi":"10.1016/j.anucene.2025.111279","DOIUrl":"10.1016/j.anucene.2025.111279","url":null,"abstract":"<div><div>Despite the proven feasibility and cost-effectiveness of nuclear desalination in small modular reactors (SMRs), the exclusive use of waste heat for this purpose remains virtually unexplored. This work investigates coupling an SMR (NuScale) to a direct contact membrane distillation (DCMD) desalination plant with heat recovery and feed recycle. Both the reactor waste heat and the low-pressure (LP) steam extraction from the reactor turbine were considered as heat sources. The DCMD hollow fiber module was modeled as a porous medium with satisfactory accuracy. Key parameters affecting system performance, identified through factorial analysis, include membrane porosity, module length, feed superficial velocity, and fiber inner radius. Single and multiobjective optimization analyses revealed the feasibility of producing up to 3,810 m<sup>3</sup>/d of water without any reactor power loss, and up to 8,832 m<sup>3</sup>/d, with a 2.28 MWe power loss, using steam extraction. DCMD demonstrated competitiveness, especially when the reactor’s primary purpose is electricity generation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111279"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520184","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 : 2025-02-28DOI: 10.1016/j.anucene.2025.111292
Evgeny Nikitin, Emil Fridman
This paper presents the development of a neutronics modeling methodology for handling non-uniform mesh deformations in a deterministic 3D reactor core dynamics simulator. Coordinate transformation is applied to solve physical problems in deformed geometries, while using a numerical solver originally derived for regular geometries. Specifically, this paper introduces and evaluates a method based on coordinate transformation to simulate non-uniform deformations of nuclear reactor cores and their impact on the reactor physics behavior.
The method, developed for the 3D reactor core simulator DYN3D, is applicable to any nodal diffusion-based reactor physics solver. It was tested in four stages using sodium-cooled fast reactors: (1) proof of concept on a simplified core; (2) feasibility of modelling flowering scenarios in a realistic core; (3) performance evaluation through the Phénix reactor core flowering experiment; and (4) applicability to dynamic deformation scenarios, such as pressure wave propagation.
{"title":"A coordinate transformation method to simulate non-uniform radial deformation of nuclear reactor cores","authors":"Evgeny Nikitin, Emil Fridman","doi":"10.1016/j.anucene.2025.111292","DOIUrl":"10.1016/j.anucene.2025.111292","url":null,"abstract":"<div><div>This paper presents the development of a neutronics modeling methodology for handling non-uniform mesh deformations in a deterministic 3D reactor core dynamics simulator. Coordinate transformation is applied to solve physical problems in deformed geometries, while using a numerical solver originally derived for regular geometries. Specifically, this paper introduces and evaluates a method based on coordinate transformation to simulate non-uniform deformations of nuclear reactor cores and their impact on the reactor physics behavior.</div><div>The method, developed for the 3D reactor core simulator DYN3D, is applicable to any nodal diffusion-based reactor physics solver. It was tested in four stages using sodium-cooled fast reactors: (1) proof of concept on a simplified core; (2) feasibility of modelling flowering scenarios in a realistic core; (3) performance evaluation through the Phénix reactor core flowering experiment; and (4) applicability to dynamic deformation scenarios, such as pressure wave propagation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111292"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520134","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 : 2025-02-28DOI: 10.1016/j.anucene.2025.111305
Ye Yang , Qian Ye , Mengyan Hu , Xueyan Zhang , Jun Yang
The Code Scaling, applicability, and Uncertainty method states the code’s capability to scale up processes from test facilities to full-scale nuclear power plants needs to be validated and evaluated. The reason for this validation is that it is infeasible (or cost prohibitive) to perform meaningful experiments at full scale and the ability of numerical tools designed to simulate the performance of nuclear reactors can be proven only at reduced scale.
ACME is an integral test facility, which is designed to study the behavior of China Advanced Pressurized Water Reactor (PWR) under accident conditions. The RELAP5 code is the best estimate thermal hydraulic system code for performing nuclear power plant safety analysis. This study validates the code scale up capability for application on China Advanced PWR. Firstly, we propose a new evaluation scheme, which is to take the realistically constructed test facility as a reference and scale up its numerical model to the size of a prototype power plant strictly according to scaling laws. This method, on one hand, ensures that the numerical model of the test facility and the scale up numerical model maintain consistency in node division. On the other hand, it avoids the influence of engineering deviations. Secondly, a numerical model for the prototype power plant scale was established based on the ideal scaling laws. After that, a 2-inch cold leg break accident test was simulated on two different scale numerical models, and the calculation results were compared with experimental data. The RELAP5 scale up capability to predict the accident phenomenon of China Advanced Gen-III PWR was evaluated using both qualitative and Fast Fourier Transform Based Method (FFTBM) quantitative methods.
{"title":"Evaluation of scale-up capability of best estimate code application on China advanced Gen-III reactor","authors":"Ye Yang , Qian Ye , Mengyan Hu , Xueyan Zhang , Jun Yang","doi":"10.1016/j.anucene.2025.111305","DOIUrl":"10.1016/j.anucene.2025.111305","url":null,"abstract":"<div><div>The Code Scaling, applicability, and Uncertainty method states the code’s capability to scale up processes from test facilities to full-scale nuclear power plants needs to be validated and evaluated. The reason for this validation is that it is infeasible (or cost prohibitive) to perform meaningful experiments at full scale and the ability of numerical tools designed to simulate the performance of nuclear reactors can be proven only at reduced scale.</div><div>ACME is an integral test facility, which is designed to study the behavior of China Advanced Pressurized Water Reactor (PWR) under accident conditions. The RELAP5 code is the best estimate thermal hydraulic system code for performing nuclear power plant safety analysis. This study validates the code scale up capability for application on China Advanced PWR. Firstly, we propose a new evaluation scheme, which is to take the realistically constructed test facility as a reference and scale up its numerical model to the size of a prototype power plant strictly according to scaling laws. This method, on one hand, ensures that the numerical model of the test facility and the scale up numerical model maintain consistency in node division. On the other hand, it avoids the influence of engineering deviations. Secondly, a numerical model for the prototype power plant scale was established based on the ideal scaling laws. After that, a 2-inch cold leg break accident test was simulated on two different scale numerical models, and the calculation results were compared with experimental data. The RELAP5 scale up capability to predict the accident phenomenon of China Advanced Gen-III PWR was evaluated using both qualitative and Fast Fourier Transform Based Method (FFTBM) quantitative methods.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111305"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520182","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 : 2025-02-28DOI: 10.1016/j.anucene.2025.111302
Muhammad Zubair, Ronak Shahin Radkiany, Mahra Mohammad Aljaradi, Shamma Abdulla Alblooshi, Horeya Sultan Alhammadi, Yumna Akram
This paper builds upon the initial Level 1 Probabilistic Safety Assessment (PSA) for Emergency Diesel Generator (EDG) failures in the APR-1400 Nuclear Power Plant (NPP). By incorporating more complex EDG failure scenarios and expanding the scope of fault trees and event trees, it provides a more comprehensive safety performance analysis using AIMS-PSA software. The results highlight the reactor’s safety systems’ resilience in the face of an increasing number of malfunctioning EDGs. An in-depth understanding of the interdependencies and failure probability of the Essential Chilled Water System (ECWS), Component Cooling Water System (CCWS), 480 V bus, and fault trees for the 125 V and 4.16 kV buses has been obtained through this detailed examination. The Core Damage Frequency (CDF) values from the analysis, 1.71E-6 /RY, were found to be significantly lower compared to official literature values of 2.56E-6 /RY and 2.25E-6 /RY, attributed to the exclusion of more extensive operational and environmental aspects of LOOP and common cause failure events. The findings provide critical insights for enhancing nuclear safety frameworks and underscore the need for scenario-based PSAs to address potential vulnerabilities. Overall, the paper not only effectively measures the resilience of the NPP’s safety measures to various failure modes but also provides regulators and plant management with insightful information about the significance of redundancy and diversity in preserving nuclear safety over the NPP’s operational lifetime.
{"title":"PSA level 1 analysis by considering redundancy and diversity of emergency diesel generators for APR-1400","authors":"Muhammad Zubair, Ronak Shahin Radkiany, Mahra Mohammad Aljaradi, Shamma Abdulla Alblooshi, Horeya Sultan Alhammadi, Yumna Akram","doi":"10.1016/j.anucene.2025.111302","DOIUrl":"10.1016/j.anucene.2025.111302","url":null,"abstract":"<div><div>This paper builds upon the initial Level 1 Probabilistic Safety Assessment (PSA) for Emergency Diesel Generator (EDG) failures in the APR-1400 Nuclear Power Plant (NPP). By incorporating more complex EDG failure scenarios and expanding the scope of fault trees and event trees, it provides a more comprehensive safety performance analysis using AIMS-PSA software. The results highlight the reactor’s safety systems’ resilience in the face of an increasing number of malfunctioning EDGs. An in-depth understanding of the interdependencies and failure probability of the Essential Chilled Water System (ECWS), Component Cooling Water System (CCWS), 480 V bus, and fault trees for the 125 V and 4.16 kV buses has been obtained through this detailed examination. The Core Damage Frequency (CDF) values from the analysis, 1.71E-6 /RY, were found to be significantly lower compared to official literature values of 2.56E-6 /RY and 2.25E-6 /RY, attributed to the exclusion of more extensive operational and environmental aspects of LOOP and common cause failure events. The findings provide critical insights for enhancing nuclear safety frameworks and underscore the need for scenario-based PSAs to address potential vulnerabilities. Overall, the paper not only effectively measures the resilience of the NPP’s safety measures to various failure modes but also provides regulators and plant management with insightful information about the significance of redundancy and diversity in preserving nuclear safety over the NPP’s operational lifetime.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111302"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520183","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 : 2025-02-28DOI: 10.1016/j.anucene.2025.111321
Mohamed Bakr Mohamed , A.M. El-naggar , Zein K. Heiba , A.M. Kamal
The present study seeks to synthesize CdS and CdS/M (M = Fe, Mg, Mn) nanoparticles incorporated into a polyvinyl alcohol (PVA)/PEG (polyethylene glycol) blend polymer film, with the objective of developing an advanced nanocomposite for diverse applications in energy storage and radiation shielding. The phases formed in the fillers samples were identified using the X-ray diffraction technique. The structural, morphology, dielectric, and radiation shielding features of the films containing the nanofiller were assessed. Doped blend with CdS/Fe demonstrated the highest dielectric constant and ac conductivity values. The doped blend including CdS has the greatest energy density values. The relaxation time affected by the kind of filler. The influence of type of filler in host blended polymer on linear attenuation coefficients (LAC), mean free path (MFP), mass attenuation coefficient (MAC), half value length (HVL), tenth value length (TVL), equivalent atomic number (Zeq), effective atomic number (Zeff), electron density (Neff), atomic cross-section (ACS), and electronic cross-section (ECS), fast neutron removal cross-section (FNRCS), buildup factor values was examined using Phy-X/PSD program. The obtained results demonstrated that doped blends displayed a propensity to interact with gamma rays instead of merely transmitting through them. Our doped blends exhibit superior neutron shielding properties compared to the undoped PVA/PEG blended polymer. The results indicated that doped blends may serve as viable and interesting nanocomposites for radiation shielding application. Also, CdS/Fe sample can be used in the energy storage capacitance application.
{"title":"Effect of CdS/M doping on the dielectric and radiation shielding of PVA/PEG composite film","authors":"Mohamed Bakr Mohamed , A.M. El-naggar , Zein K. Heiba , A.M. Kamal","doi":"10.1016/j.anucene.2025.111321","DOIUrl":"10.1016/j.anucene.2025.111321","url":null,"abstract":"<div><div>The present study seeks to synthesize CdS and CdS/M (<em>M</em> = Fe, Mg, Mn) nanoparticles incorporated into a polyvinyl alcohol (PVA)/PEG (polyethylene glycol) blend polymer film, with the objective of developing an advanced nanocomposite for diverse applications in energy storage and radiation shielding. The phases formed in the fillers samples were identified using the X-ray diffraction technique. The structural, morphology, dielectric, and radiation shielding features of the films containing the nanofiller were assessed. Doped blend with CdS/Fe demonstrated the highest dielectric constant and ac conductivity values. The doped blend including CdS has the greatest energy density values. The relaxation time affected by the kind of filler. The influence of type of filler in host blended polymer on linear attenuation coefficients (LAC), mean free path (MFP), mass attenuation coefficient (MAC), half value length (HVL), tenth value length (TVL), equivalent atomic number (<em>Z</em><sub>eq</sub>), effective atomic number (<em>Z<sub>e</sub></em><sub>ff</sub>), electron density (<em>N<sub>eff</sub>)</em>, atomic cross-section (ACS), and electronic cross-section (ECS), fast neutron removal cross-section (FNRCS), buildup factor values was examined using Phy-X/PSD program. The obtained results demonstrated that doped blends displayed a propensity to interact with gamma rays instead of merely transmitting through them. Our doped blends exhibit superior neutron shielding properties compared to the undoped PVA/PEG blended polymer. The results indicated that doped blends may serve as viable and interesting nanocomposites for radiation shielding application. Also, CdS/Fe sample can be used in the energy storage capacitance application.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111321"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520181","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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