Nuclear Engineering and Design最新文献

英文中文

Effects of second phase particles and pores on grain boundary migration during solid-state sintering: A phase-field study

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113800

Xiaoyong Qi , Yanbo Jiang , Wenlong Shen , Jinli Cao , Yuxuan Liao , Wenbo Liu

In the present work, a new phase-field model was developed to simulate the effects of second phase particles and pores on grain boundary (GB) removement during solid-state sintering of ceramic nuclear fuel. The simulation results show that the second phase particles near the GB make the interface energy distribution steeper, and the GB-particle equilibrium dihedral angle is related to the GB-particle interface energy and the GB-pore interface energy. During sintering, the contact between particles and GB in the early stage promotes the shrinkage of GB. However, during the later stage, the shrinkage of GB is suppressed due to the interaction of second phase particles and crystal grains. When the particles and pores act on the GB at the same time, the GB usually move together with the pores, while the GB prefer to leave the particles ultimately. The sintering simulation of polycrystalline containing both second-phase particles and pores showed that, when the particle area fraction increases, the growth rate of crystal grains slows down and the final grain size decreases. Consequently, second phase particles with different volume fractions can be doped to control the final average grain size during sintering.

{"title":"Effects of second phase particles and pores on grain boundary migration during solid-state sintering: A phase-field study","authors":"Xiaoyong Qi , Yanbo Jiang , Wenlong Shen , Jinli Cao , Yuxuan Liao , Wenbo Liu","doi":"10.1016/j.nucengdes.2024.113800","DOIUrl":"10.1016/j.nucengdes.2024.113800","url":null,"abstract":"<div><div>In the present work, a new phase-field model was developed to simulate the effects of second phase particles and pores on grain boundary (GB) removement during solid-state sintering of ceramic nuclear fuel. The simulation results show that the second phase particles near the GB make the interface energy distribution steeper, and the GB-particle equilibrium dihedral angle is related to the GB-particle interface energy and the GB-pore interface energy. During sintering, the contact between particles and GB in the early stage promotes the shrinkage of GB. However, during the later stage, the shrinkage of GB is suppressed due to the interaction of second phase particles and crystal grains. When the particles and pores act on the GB at the same time, the GB usually move together with the pores, while the GB prefer to leave the particles ultimately. The sintering simulation of polycrystalline containing both second-phase particles and pores showed that, when the particle area fraction increases, the growth rate of crystal grains slows down and the final grain size decreases. Consequently, second phase particles with different volume fractions can be doped to control the final average grain size during sintering.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113800"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167571","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}

引用次数: 0

Evaluation of Cs+ and Sr2+ sorption behaviour of synthetic NaA and NaP zeolite modified with graphene oxide

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113756

Y. Raghavendra , Santanu Bera , T.V. Krishna Mohan , S.N. Achary

NaP and NaA zeolites are the promising materials for removal of cesium (Cs⁺) and strontium (Sr²⁺) from nuclear waste, and the effective utilisation of these relies on the appropriate modification to enhance their ion exchange properties. In this aim, zeolite (NaZ) containing both NaA and NaP zeolites, and the composite of zeolite and graphene oxide (GO-NaZ), were synthesised by hydrothermal process and characterized by varieties of techniques, like XRD, Raman spectroscopy, scanning electron microscopy (SEM), gas adsorption, and small angle X-ray scattering (SAXS). Presence of both NaP and NaA phases was confirmed from XRD and Raman spectroscopy. SEM, SAXS and gas adsorption studies indicated presence of extensive macropores and rough surfaces in the samples. The sorption behaviours of the samples were studied using solutions of radioactive tracer or inactive ions, and the sorption efficiencies were evaluated by equilibrium and kinetic studies. Equilibrium studies indicated that the composites have higher Cs⁺ exchange capacity than the pristine NaZ. However, opposite trend was observed for Sr²⁺ ions. Kinetic studies revealed that the Sr²⁺ and Cs⁺ uptake follow pseudo-2nd order kinetic model. It was also revealed that GO-NaZ composite shows better selectivity for Cs⁺ over Na⁺ at lower concertation of Cs⁺, and hence it can be a promising material for site decontamination.

{"title":"Evaluation of Cs+ and Sr2+ sorption behaviour of synthetic NaA and NaP zeolite modified with graphene oxide","authors":"Y. Raghavendra , Santanu Bera , T.V. Krishna Mohan , S.N. Achary","doi":"10.1016/j.nucengdes.2024.113756","DOIUrl":"10.1016/j.nucengdes.2024.113756","url":null,"abstract":"<div><div>NaP and NaA zeolites are the promising materials for removal of cesium (Cs<sup>+</sup>) and strontium (Sr<sup>2+</sup>) from nuclear waste, and the effective utilisation of these relies on the appropriate modification to enhance their ion exchange properties. In this aim, zeolite (NaZ) containing both NaA and NaP zeolites, and the composite of zeolite and graphene oxide (GO-NaZ), were synthesised by hydrothermal process and characterized by varieties of techniques, like XRD, Raman spectroscopy, scanning electron microscopy (SEM), gas adsorption, and small angle X-ray scattering (SAXS). Presence of both NaP and NaA phases was confirmed from XRD and Raman spectroscopy. SEM, SAXS and gas adsorption studies indicated presence of extensive macropores and rough surfaces in the samples. The sorption behaviours of the samples were studied using solutions of radioactive tracer or inactive ions, and the sorption efficiencies were evaluated by equilibrium and kinetic studies. Equilibrium studies indicated that the composites have higher Cs<sup>+</sup> exchange capacity than the pristine NaZ. However, opposite trend was observed for Sr<sup>2+</sup> ions. Kinetic studies revealed that the Sr<sup>2+</sup> and Cs<sup>+</sup> uptake follow pseudo-2nd order kinetic model. It was also revealed that GO-NaZ composite shows better selectivity for Cs<sup>+</sup> over Na<sup>+</sup> at lower concertation of Cs<sup>+</sup>, and hence it can be a promising material for site decontamination.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113756"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167572","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}

引用次数: 0

Multi-scale contact characteristics and leakage prediction of flange seal based on fractal geometry

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2025.113835

Feng Li , Lushuai Xu , Shaohua Dong , Dongying Wang , Xiujuan Dong , Biao Pan , Quan Liu

Flanges serve as the sealing and connecting components in nuclear power plant pipelines and pressure vessels. The contact gaps between flanges and gaskets can lead to flange sealing failure. Therefore, optimal flange sealing surfaces are essential for ensuring the safety and reliability of airtight pipeline systems. This paper examined the flange sealing surfaces in a nuclear power plant. A sealing contact model was established according to the characteristic parameters of the surface contours measured during the experiment, while the finite element method was used to determine the sealing contact characteristics and predict the leakage of the rough sealing contact surface. The results indicated that the surface roughness had a multi-scale effect while the sealing state was related to the sealing characteristics scale. The sealing surface morphology, material properties, and operational conditions directly affected the sealing interface structure and sealing performance. In addition, the contact clearance of the sealing surface decreased in conjunction with increased applied load, normal compression displacement, and surface roughness, while the surface leakage rate increased at a higher surface contact clearance and pipeline conveying pressure.

{"title":"Multi-scale contact characteristics and leakage prediction of flange seal based on fractal geometry","authors":"Feng Li , Lushuai Xu , Shaohua Dong , Dongying Wang , Xiujuan Dong , Biao Pan , Quan Liu","doi":"10.1016/j.nucengdes.2025.113835","DOIUrl":"10.1016/j.nucengdes.2025.113835","url":null,"abstract":"<div><div>Flanges serve as the sealing and connecting components in nuclear power plant pipelines and pressure vessels. The contact gaps between flanges and gaskets can lead to flange sealing failure. Therefore, optimal flange sealing surfaces are essential for ensuring the safety and reliability of airtight pipeline systems. This paper examined the flange sealing surfaces in a nuclear power plant. A sealing contact model was established according to the characteristic parameters of the surface contours measured during the experiment, while the finite element method was used to determine the sealing contact characteristics and predict the leakage of the rough sealing contact surface. The results indicated that the surface roughness had a multi-scale effect while the sealing state was related to the sealing characteristics scale. The sealing surface morphology, material properties, and operational conditions directly affected the sealing interface structure and sealing performance. In addition, the contact clearance of the sealing surface decreased in conjunction with increased applied load, normal compression displacement, and surface roughness, while the surface leakage rate increased at a higher surface contact clearance and pipeline conveying pressure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113835"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168318","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}

引用次数: 0

Analysis of TRISO failure fraction in PeLUIt reactor with increasing power

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2025.113842

Fitria Miftasani , Steven Wijaya , Nina Widiawati , Anni Nuril Hidayati , Dany Mulyana , Hakimul Wafda , Anik Purwaningsih , Fajar Al Afghani , Muhammad Ilham Bayquni , Arya Adhyaksa Waskita , Topan Setiadipura

This study investigates the impact of increasing power in the PeLUIt reactor on the failure fraction of TRISO-coated fuel particles during both normal operation and Depressurized Loss of Forced Cooling (DLOFC) accident scenarios. In this study, the core geometry is preserved despite of the power changes while the coolant velocity is adjusted to maintain the intlet and outlet temperatures. The TRIAC-BATAN code was used to evalute the failure fraction of TRISO-coated fuel particles at various reactor power ranging from 10 MWt to 50 MWt. Neutronic calculations for PeLUIt were conducted using the PEBBED code, while irradiation and DLOFC accident temperatures were analyzed through a combination of PEBBED 1-D and THERMIX-KONVEX, coupled with PEBBED. The results show that at lower power (10 MWt to 30 MWt), the failure fraction of TRISO particles remains low, with minimal increases as power rise. However, above 30 MWt, an exponential increase in failure fraction is observed, particularly beyond 40 MWt. At 50 MWt, when fuel temperatures exceed the safety threshold of 1600 °C during DLOFC, a significant rise in TRISO particle failure occurs, accompanied by degradation of the SiC layer.

{"title":"Analysis of TRISO failure fraction in PeLUIt reactor with increasing power","authors":"Fitria Miftasani , Steven Wijaya , Nina Widiawati , Anni Nuril Hidayati , Dany Mulyana , Hakimul Wafda , Anik Purwaningsih , Fajar Al Afghani , Muhammad Ilham Bayquni , Arya Adhyaksa Waskita , Topan Setiadipura","doi":"10.1016/j.nucengdes.2025.113842","DOIUrl":"10.1016/j.nucengdes.2025.113842","url":null,"abstract":"<div><div>This study investigates the impact of increasing power in the PeLUIt reactor on the failure fraction of TRISO-coated fuel particles during both normal operation and Depressurized Loss of Forced Cooling (DLOFC) accident scenarios. In this study, the core geometry is preserved despite of the power changes while the coolant velocity is adjusted to maintain the intlet and outlet temperatures. The TRIAC-BATAN code was used to evalute the failure fraction of TRISO-coated fuel particles at various reactor power ranging from 10 MWt to 50 MWt. Neutronic calculations for PeLUIt were conducted using the PEBBED code, while irradiation and DLOFC accident temperatures were analyzed through a combination of PEBBED 1-D and THERMIX-KONVEX, coupled with PEBBED. The results show that at lower power (10 MWt to 30 MWt), the failure fraction of TRISO particles remains low, with minimal increases as power rise. However, above 30 MWt, an exponential increase in failure fraction is observed, particularly beyond 40 MWt. At 50 MWt, when fuel temperatures exceed the safety threshold of 1600 °C during DLOFC, a significant rise in TRISO particle failure occurs, accompanied by degradation of the SiC layer.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113842"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168319","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}

引用次数: 0

Probabilistic burst pressure estimation for steam generator tubes with fretting wear

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113789

Daeyeop Kwon , Heejae Shin , Young-Jin Oh , Kuk-Hee Lee , Chi Bum Bahn

Considering steam generator (SG) tubes form the boundaries between the primary and secondary sides of nuclear power plants, maintaining their structural integrity, even in the presence of defects in tubes, is crucial. In this study, six burst pressure models were compared using available burst pressure test data for SG tubes with wear defects. The model that accounted for both the maximum and minimum depths of wear defects exhibited the best performance in predicting the burst pressure of flat and tapered wear defects. With the most suitable burst pressure model selected, probabilistic methods were applied to evaluate the structural integrity of SG tubes with wear defects. The proposed defect repair criteria, which consider inspection intervals of 1.5 or 3.0 effective full power years and a 5% burst probability, resulted in a less conservative evaluation compared to the existing repair criteria (40% through-wall).

引用次数: 0

Review on flow and heat transfer processes in rod bundle channels of water-cooled nuclear reactor

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113799

Lipeng Du , Xiang Chen , Wenchao Zhang , Jianchuang Sun , Weihua Cai

The thermal–hydraulic analysis of coolant flow in the reactor core plays a significant role in the optimized design of nuclear fuel assemblies and nuclear safety. This article reviews the experimental and Computational Fluid Dynamics (CFD) methods for the thermal–hydraulic characteristics of subchannels with water as the coolant in different rod bundle geometries and flow conditions over the past few decades. It summarizes the effects of design parameters such as rod spacing and spacer grid type, as well as flow parameters like pressure, mass flow rate, and heat flux on fluid flow and heat transfer. Additionally, various models used for numerical simulations of rod bundle channels are introduced, providing valuable references for research and practice in related fields.

引用次数: 0

Analysis of Thermalhydraulics characteristics on constant and variable models for pressure tube radial expansion by using the ASSERT code in a CANDU6 channel

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113823

Eun Hyun Ryu

Because of extreme conditions such as high temperature, high pressure and irradiation, gradual radial expansion of pressure tube in the CANadian Deuterium and Uranium 6 (CANDU6) reactors occurs in the radial direction. Due to the decrease in the thermal margin of the channel, the Trip Set Point (TSP) for reactor trip should be decreased as well, to ensure a sufficient magnitude of safety margin for the reactor. Then operating power should be reduced as time goes by. Making precise thermal margin predictions is therefore a crucial activity. The ASSERT code has sub-channel level resolution and 2-dimensional analysis is possible. For these reasons, a lot of activities are performed using the ASSERT code. In this study, the ASSERT code calculation results from different pressure tube radial expansion models were compared with each other. With this comparison, it was examined how details in the model affect the critical results, such as Critical Channel Power (CCP) and T/H characteristics. One model used a constant radial expansion quantity along the axial flow direction, and the other using varying radial expansion quantity along the axial flow direction. The expansion area or volume was conserved so that consistency was maintained in this study.

{"title":"Analysis of Thermalhydraulics characteristics on constant and variable models for pressure tube radial expansion by using the ASSERT code in a CANDU6 channel","authors":"Eun Hyun Ryu","doi":"10.1016/j.nucengdes.2024.113823","DOIUrl":"10.1016/j.nucengdes.2024.113823","url":null,"abstract":"<div><div>Because of extreme conditions such as high temperature, high pressure and irradiation, gradual radial expansion of pressure tube in the CANadian Deuterium and Uranium 6 (CANDU6) reactors occurs in the radial direction. Due to the decrease in the thermal margin of the channel, the Trip Set Point (TSP) for reactor trip should be decreased as well, to ensure a sufficient magnitude of safety margin for the reactor. Then operating power should be reduced as time goes by. Making precise thermal margin predictions is therefore a crucial activity. The ASSERT code has sub-channel level resolution and 2-dimensional analysis is possible. For these reasons, a lot of activities are performed using the ASSERT code. In this study, the ASSERT code calculation results from different pressure tube radial expansion models were compared with each other. With this comparison, it was examined how details in the model affect the critical results, such as Critical Channel Power (CCP) and T/H characteristics. One model used a constant radial expansion quantity along the axial flow direction, and the other using varying radial expansion quantity along the axial flow direction. The expansion area or volume was conserved so that consistency was maintained in this study.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113823"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168329","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}

引用次数: 0

Numerical simulation of seismic response of spent fuel pool considering structure-fluid-structure interaction

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2025.113892

Tsvetan Genov, Joao Travanca, Stoyan Andreev, Anton Andonov

The objective of the paper is to assess approaches to include the effect of fluid–structure along with equipment-fluid–structure interaction by evaluating the response of a typical spent fuel pool found in Westinghouse type Pressurized water reactor auxiliary building to earthquake loading, as well as to derive the seismic demand in critical structural members for the needs of a seismic fragility analysis. Different approaches are presented to account for fluid–structure interaction during the dynamic nature of seismic loading. All structural key aspects considered relevant are included in the analyses: the concrete box structure, the outlining stainless-steel liner, its connections, the submerged rack’s structure, and the containing fluid.

Three numerical models were developed and tested with the finite element code LS-Dyna with different implementations considering the fluid–structure interaction to demonstrate the robustness of the structure against earthquakes. Mechanical analogue adopted in codes with spring-mass system representation of the fluid attached to the structure and explicit modelling of the domain representing the water considering the two-way interaction between structure and fluid.

It is of great importance to include all the effects of fluid–structure interaction, as they increase the response and load of the structure. Both approaches considered for fluid representation were found to capture accurately the global response of the structure. The lattice structure of the racks exerts extra damping in the fluid motion and the pressure loading on the tank walls in near vicinity during earthquake excitation. The explicit modelling approach for the fluid elements could possibly result in lowering the racks critical displacements.

{"title":"Numerical simulation of seismic response of spent fuel pool considering structure-fluid-structure interaction","authors":"Tsvetan Genov, Joao Travanca, Stoyan Andreev, Anton Andonov","doi":"10.1016/j.nucengdes.2025.113892","DOIUrl":"10.1016/j.nucengdes.2025.113892","url":null,"abstract":"<div><div>The objective of the paper is to assess approaches to include the effect of fluid–structure along with equipment-fluid–structure interaction by evaluating the response of a typical spent fuel pool found in Westinghouse type Pressurized water reactor auxiliary building to earthquake loading, as well as to derive the seismic demand in critical structural members for the needs of a seismic fragility analysis. Different approaches are presented to account for fluid–structure interaction during the dynamic nature of seismic loading. All structural key aspects considered relevant are included in the analyses: the concrete box structure, the outlining stainless-steel liner, its connections, the submerged rack’s structure, and the containing fluid.</div><div>Three numerical models were developed and tested with the finite element code LS-Dyna with different implementations considering the fluid–structure interaction to demonstrate the robustness of the structure against earthquakes. Mechanical analogue adopted in codes with spring-mass system representation of the fluid attached to the structure and explicit modelling of the domain representing the water considering the two-way interaction between structure and fluid.</div><div>It is of great importance to include all the effects of fluid–structure interaction, as they increase the response and load of the structure. Both approaches considered for fluid representation were found to capture accurately the global response of the structure. The lattice structure of the racks exerts extra damping in the fluid motion and the pressure loading on the tank walls in near vicinity during earthquake excitation. The explicit modelling approach for the fluid elements could possibly result in lowering the racks critical displacements.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113892"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169293","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}

引用次数: 0

Investigation on the effect of operational parameters in a microreactor system on the morphology and size distribution of thorium oxalate

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113724

P. Zaheri, S. Ammari Allahyari, A. charkhi

Thorium oxide has recently garnered attention as a nuclear fuel due to the scarcity of uranium resources and the abundance of thorium resources, as well as its favorable thermal and neutronic properties. One of the most desirable characteristics of thorium nuclear fuels is their thermal properties, which are affected by the size distribution and morphology of thorium oxide particles. Thorium nitrate is the most common method for producing thorium oxide. Since no purification processes occur in the production of thorium oxide from oxalates, particle size and shape control are crucial. In this paper, the synthesis of thorium oxalate particles is carried out in a microreactor system to control the parameters of the precipitates. The main parameters that influence the efficiency, size distribution, and morphology of thorium oxalate are the concentration ratio of oxalic acid to thorium nitrate solution, as well as the flow rate ratio of these feed materials. Results show that at lower flow rate ratios of thorium nitrate to oxalic acid solution and higher concentration ratios of acid to thorium nitrate solution, a uniform particle size distribution and smaller particles are obtained, which are suitable for further calcination to prepare high-density and small grain size pellets.

{"title":"Investigation on the effect of operational parameters in a microreactor system on the morphology and size distribution of thorium oxalate","authors":"P. Zaheri, S. Ammari Allahyari, A. charkhi","doi":"10.1016/j.nucengdes.2024.113724","DOIUrl":"10.1016/j.nucengdes.2024.113724","url":null,"abstract":"<div><div>Thorium oxide has recently garnered attention as a nuclear fuel due to the scarcity of uranium resources and the abundance of thorium resources, as well as its favorable thermal and neutronic properties. One of the most desirable characteristics of thorium nuclear fuels is their thermal properties, which are affected by the size distribution and morphology of thorium oxide particles. Thorium nitrate is the most common method for producing thorium oxide. Since no purification processes occur in the production of thorium oxide from oxalates, particle size and shape control are crucial. In this paper, the synthesis of thorium oxalate particles is carried out in a microreactor system to control the parameters of the precipitates. The main parameters that influence the efficiency, size distribution, and morphology of thorium oxalate are the concentration ratio of oxalic acid to thorium nitrate solution, as well as the flow rate ratio of these feed materials. Results show that at lower flow rate ratios of thorium nitrate to oxalic acid solution and higher concentration ratios of acid to thorium nitrate solution, a uniform particle size distribution and smaller particles are obtained, which are suitable for further calcination to prepare high-density and small grain size pellets.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113724"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167567","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}

引用次数: 0

Finite element formulations of the semi-analytical time-domain model for flow-induced vibration of tube bundles in steam generators

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-01 DOI: 10.1016/j.nucengdes.2024.113769

Pan Sun , Junzhe Shen , Xielin Zhao , Jinxiong Zhou

This paper describes the finite element method (FEM) formulations for flow-induced vibration of steam generator tube bundles, based on a semi-analytical time domain (SATD) model proposed by us recently (Sun et al., Appl. Math. Modell. 132, 2024, 252-273). The SATD model is featured by explicit fluid force expressions in terms of multiple integrals with time delays. The new SATD model allows for a unified treatment of both time domain and frequency domain fluidelastic instability (FEI) analysis of a tube due subjected to crossflow. Specifically, this study adds to the body of knowledge through two contributions: (1) This paper presents the details of FEM formulations on this newly formulated SATD theory, in particular the discretization of fluid forces in both lift and drag directions, which are crucial for FEM implementation but missing in the literature; (2) This study presents the formulations based on FEM discretization for frequency domain FEI analysis of tubes, and provides very the first reference on FEI stability analysis of U-tubes. Our FEM results agree well with reported experimental data and numerical results. The numerical examples presented here, including single-span straight tubes subjected to uniform flow and multi-span U-tubes subjected to nonuniform flow, not only demonstrate the merit of the theory for academic research, but also exhibit the potential of the FEM code for realistic engineering applications. Our efforts provide useful and powerful numerical tools for flow-induced vibration analysis of tube bundles in steam generators and other heat exchangers.

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