Performance of two distinct fuel systems, U-7Mo fuel in zircaloy cladding and U-10Mo fuel in aluminum cladding was studied. First, a mini plate with zircaloy cladding from a previous experiment was evaluated via finite element analysis. By using the same plate geometry and irradiation conditions, another plate consisting of U-10Mo fuel and Al6061-O cladding was simulated. The results were then comparatively evaluated to explore the feasibility of employing zircaloy as an alternative cladding. Simulations indicated the zircaloy cladding plate would operate roughly 50°C hotter as compared with the Al alloy cladding plate. Larger deformations in the thickness direction for the plate with zircaloy cladding were noted. It was observed that the post-fabrication stresses in the fuel would be relieved quickly in the reactor, regardless of cladding type. Although the fuel stresses would still develop at reactor shutdown, the fuel would be stress-free during the irradiation for both cladding types. At shutdown, the plate with zircaloy cladding would have higher stresses due to higher operating temperatures. Similarly, the stresses after shutdown are higher in the foil core for the plates with zircaloy cladding. The Al cladding plate would have higher plastic strains as compared with the zircaloy cladding plate. The zircaloy cladding plate is significantly stiffer, causing higher stresses in the fuel and at the interface. Overall, employing zircaloy as an alternate cladding is not expected to produce a more favorable thermomechanical performance as compared to the performance of an Al alloy cladding plate.
{"title":"Thermomechanical Performance Assessment of U-Mo Monolithic Fuel Plates with Zircaloy Cladding","authors":"H. Ozaltun, J. Cole, B. Rabin","doi":"10.1115/1.4063163","DOIUrl":"https://doi.org/10.1115/1.4063163","url":null,"abstract":"\u0000 Performance of two distinct fuel systems, U-7Mo fuel in zircaloy cladding and U-10Mo fuel in aluminum cladding was studied. First, a mini plate with zircaloy cladding from a previous experiment was evaluated via finite element analysis. By using the same plate geometry and irradiation conditions, another plate consisting of U-10Mo fuel and Al6061-O cladding was simulated. The results were then comparatively evaluated to explore the feasibility of employing zircaloy as an alternative cladding. Simulations indicated the zircaloy cladding plate would operate roughly 50°C hotter as compared with the Al alloy cladding plate. Larger deformations in the thickness direction for the plate with zircaloy cladding were noted. It was observed that the post-fabrication stresses in the fuel would be relieved quickly in the reactor, regardless of cladding type. Although the fuel stresses would still develop at reactor shutdown, the fuel would be stress-free during the irradiation for both cladding types. At shutdown, the plate with zircaloy cladding would have higher stresses due to higher operating temperatures. Similarly, the stresses after shutdown are higher in the foil core for the plates with zircaloy cladding. The Al cladding plate would have higher plastic strains as compared with the zircaloy cladding plate. The zircaloy cladding plate is significantly stiffer, causing higher stresses in the fuel and at the interface. Overall, employing zircaloy as an alternate cladding is not expected to produce a more favorable thermomechanical performance as compared to the performance of an Al alloy cladding plate.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"62 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90633481","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
To achieve the long-term challenge of nuclear energy public acceptance in Thailand, nuclear safety research needed to be properly determined in both domestic and international directions, especially in the 2020s which was a period passing the Fukushima disaster over 10 years. Thailand Institute of Nuclear Technology (TINT) has studied nuclear safety research after the Fukushima accident to answer technical and social issues of nuclear power. An update of nuclear safety research from domestic experts and international surveys was needed in order to identify potential collaborative research to serve the goal of public acceptance reinforcement. The objective of this study was to survey, assess and rank the importance and knowledge level of nuclear safety research in Thailand among domestic experts in various fields. The survey was extended to collect the opinion of international participants of the ASEAN Network on Nuclear Power Safety Research (ASEAN NPSR) to analyze the similarity of the nuclear research interest for reinforcing the future collaborative project. As a result, the importance and knowledge level showed diverse important research topics with the priority of research scopes on human factor, novel reactor technologies, and risk assessment. According to the ASEAN NPSR survey, the nuclear safety research of severe accident, risk assessment, and novel reactor technologies were listed as potential collaborative projects. Also, the domestic and ASEAN NPSR survey results helped support the new collaborative research extension session in the annual ASEAN NPSR meeting to together discuss the potential nuclear safety research between members for the 2020s.
{"title":"Survey Analysis of Potential Nuclear Safety Research of Thailand for International Research Collaborative Reinforcement in the 2020s","authors":"W. Vechgama, K. Silva","doi":"10.1115/1.4063162","DOIUrl":"https://doi.org/10.1115/1.4063162","url":null,"abstract":"\u0000 To achieve the long-term challenge of nuclear energy public acceptance in Thailand, nuclear safety research needed to be properly determined in both domestic and international directions, especially in the 2020s which was a period passing the Fukushima disaster over 10 years. Thailand Institute of Nuclear Technology (TINT) has studied nuclear safety research after the Fukushima accident to answer technical and social issues of nuclear power. An update of nuclear safety research from domestic experts and international surveys was needed in order to identify potential collaborative research to serve the goal of public acceptance reinforcement. The objective of this study was to survey, assess and rank the importance and knowledge level of nuclear safety research in Thailand among domestic experts in various fields. The survey was extended to collect the opinion of international participants of the ASEAN Network on Nuclear Power Safety Research (ASEAN NPSR) to analyze the similarity of the nuclear research interest for reinforcing the future collaborative project. As a result, the importance and knowledge level showed diverse important research topics with the priority of research scopes on human factor, novel reactor technologies, and risk assessment. According to the ASEAN NPSR survey, the nuclear safety research of severe accident, risk assessment, and novel reactor technologies were listed as potential collaborative projects. Also, the domestic and ASEAN NPSR survey results helped support the new collaborative research extension session in the annual ASEAN NPSR meeting to together discuss the potential nuclear safety research between members for the 2020s.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"9 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86769508","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
T. Beuthe, Aleksandar Vasić, C. Azih, Pablo Diazgomezmaqueo
Advanced small modular reactors strive to improve reactor safety though increased utilization of passive heat transport and safety systems. An innovative means of meeting this design goal is to use alkali metal heat pipes to cool the reactor under both normal and abnormal operating conditions. A heat pipe model has been added to the ARIANT thermalhydraulic code to enable reactor modelling and support the design and licensing of new reactors. Saturation fluid properties are a fundamental input to this model. Consequently, this article provides a comprehensive comparative overview of the best available sodium saturation property correlations developed over the past century. The results show most of the sodium property correlations needed to model a heat pipe are relatively well defined and recommendations for their use can be provided.
{"title":"Development Of Analysis Tools For Heat Pipes Used In Small Modular Reactors: Sodium Property Correlations","authors":"T. Beuthe, Aleksandar Vasić, C. Azih, Pablo Diazgomezmaqueo","doi":"10.1115/1.4063112","DOIUrl":"https://doi.org/10.1115/1.4063112","url":null,"abstract":"\u0000 Advanced small modular reactors strive to improve reactor safety though increased utilization of passive heat transport and safety systems. An innovative means of meeting this design goal is to use alkali metal heat pipes to cool the reactor under both normal and abnormal operating conditions. A heat pipe model has been added to the ARIANT thermalhydraulic code to enable reactor modelling and support the design and licensing of new reactors. Saturation fluid properties are a fundamental input to this model. Consequently, this article provides a comprehensive comparative overview of the best available sodium saturation property correlations developed over the past century. The results show most of the sodium property correlations needed to model a heat pipe are relatively well defined and recommendations for their use can be provided.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"139 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83601764","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The molten salt reactor (MSR) concept is among the Generation IV designs considered feasible for providing clean, safe, sustainable, and economical energy supplies to the world's population. The depletion of fuel for a small modular fluoride molten salt reactor (sm-FMSR) with a closed fuel cycle based on the integral molten salt reactor (IMSR) concept has been investigated using Serpent. Three fueling schemes to control Serpent depletion cycles have been simulated and compared: step fueling (SF), continuous fueling with all fission products (FPs) accumulating in the reactor system (CFA), and continuous fueling with insoluble FPs separated from fuel (CFS). Sm-FMSRs with SF and with CFA require similar quantities of "top-up" fuel, consume similar fuel (fissile) amounts, and result in similar fuel isotopic concentrations if keff is kept within a similar range. However, with separation of insoluble FPs from the circulating fuel, CFS gains a large reactivity worth due to the removal of FP poisons. This allows for reduction of fuel enrichment in both initial and total top-up fuel, and leads to savings of a considerable fissile quantity in fueling MSR and in spent fuel. The Serpent depletion calculations require manual arithmetic calculations for adjustment of the Serpent built-in settings before the start of every calculation cycle for all three fueling schemes. Implementation of additional Serpent flow features in changing material volumes and flow constants would facilitate the simulation of the fuel depletion process and allow for more realistic simulations of fuel circulation.
{"title":"Depletion Calculations For An Integral Small Molten Salt Reactor With Serpent","authors":"Xiaolin Wang, T. S. Nguyen, D. Wojtaszek","doi":"10.1115/1.4063111","DOIUrl":"https://doi.org/10.1115/1.4063111","url":null,"abstract":"\u0000 The molten salt reactor (MSR) concept is among the Generation IV designs considered feasible for providing clean, safe, sustainable, and economical energy supplies to the world's population. The depletion of fuel for a small modular fluoride molten salt reactor (sm-FMSR) with a closed fuel cycle based on the integral molten salt reactor (IMSR) concept has been investigated using Serpent. Three fueling schemes to control Serpent depletion cycles have been simulated and compared: step fueling (SF), continuous fueling with all fission products (FPs) accumulating in the reactor system (CFA), and continuous fueling with insoluble FPs separated from fuel (CFS). Sm-FMSRs with SF and with CFA require similar quantities of \"top-up\" fuel, consume similar fuel (fissile) amounts, and result in similar fuel isotopic concentrations if keff is kept within a similar range. However, with separation of insoluble FPs from the circulating fuel, CFS gains a large reactivity worth due to the removal of FP poisons. This allows for reduction of fuel enrichment in both initial and total top-up fuel, and leads to savings of a considerable fissile quantity in fueling MSR and in spent fuel. The Serpent depletion calculations require manual arithmetic calculations for adjustment of the Serpent built-in settings before the start of every calculation cycle for all three fueling schemes. Implementation of additional Serpent flow features in changing material volumes and flow constants would facilitate the simulation of the fuel depletion process and allow for more realistic simulations of fuel circulation.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"10 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88048095","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
F. Hernández, P. Arena, L. Boccaccini, I. Cristescu, A. Del Nevo, Pierre Sardain, G. Spagnuolo, M. Utili, A. Venturini, Guangming Zhou
The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021.
{"title":"Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities","authors":"F. Hernández, P. Arena, L. Boccaccini, I. Cristescu, A. Del Nevo, Pierre Sardain, G. Spagnuolo, M. Utili, A. Venturini, Guangming Zhou","doi":"10.3390/jne4030037","DOIUrl":"https://doi.org/10.3390/jne4030037","url":null,"abstract":"The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"25 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83332011","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect, but up to now, no experimental program has confirmed this effect. The French Plutonium Temperature Effect Experimental Program (or PU+ in short) aims to effectively show that such a positive temperature effect exists for diluted plutonium solutions. The PU+ experiments were conducted in the “Apparatus B” facility at the CEA VALDUC research center in France. It involved several sub-critical approach-type experiments using plutonium nitrate solutions with concentrations of 14.3, 15, and 20 g/L at temperatures ranging from 20 to 40 °C. Fourteen (five at 20 g/L, four at 15 g/L, and five at 14.3 g/L) phase I experiments (consisting of independent sub-critical approaches) were performed between 2006 and 2007. The impact of the uncertainties on solution acidity and plutonium concentration made it difficult to demonstrate the positive temperature effect, requiring an additional phase II experiment (with a unique plutonium solution) from 22 to 28 °C that was performed in July 2007. This phase II experiment has shown the existence of a positive temperature effect of ~+5.17 pcm/°C (from 22 to 28 °C for a plutonium concentration of 14.3 g/L). It has recently been possible to confirm the results of this program with MORET 5 calculations by generating thermal scattering data S(α,β) at the correct experimental temperatures. This paper finally presents a fully documented experimental program highlighting the Plutonium Temperature Effect theoretically described in the literature. Its high level of precision and its “one-step” approach to criticality allowed it to show a significant positive temperature effect for a rather small variation of temperature (+6 °C). The order of magnitude of the effect was confirmed with Monte Carlo calculations using thermal scattering data for hydrogen in the solution produced by IRSN for the purpose of the comparison.
{"title":"The Plutonium Temperature Effect Program","authors":"N. Leclaire, Vaibhav Jaiswal","doi":"10.3390/jne4030035","DOIUrl":"https://doi.org/10.3390/jne4030035","url":null,"abstract":"Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect, but up to now, no experimental program has confirmed this effect. The French Plutonium Temperature Effect Experimental Program (or PU+ in short) aims to effectively show that such a positive temperature effect exists for diluted plutonium solutions. The PU+ experiments were conducted in the “Apparatus B” facility at the CEA VALDUC research center in France. It involved several sub-critical approach-type experiments using plutonium nitrate solutions with concentrations of 14.3, 15, and 20 g/L at temperatures ranging from 20 to 40 °C. Fourteen (five at 20 g/L, four at 15 g/L, and five at 14.3 g/L) phase I experiments (consisting of independent sub-critical approaches) were performed between 2006 and 2007. The impact of the uncertainties on solution acidity and plutonium concentration made it difficult to demonstrate the positive temperature effect, requiring an additional phase II experiment (with a unique plutonium solution) from 22 to 28 °C that was performed in July 2007. This phase II experiment has shown the existence of a positive temperature effect of ~+5.17 pcm/°C (from 22 to 28 °C for a plutonium concentration of 14.3 g/L). It has recently been possible to confirm the results of this program with MORET 5 calculations by generating thermal scattering data S(α,β) at the correct experimental temperatures. This paper finally presents a fully documented experimental program highlighting the Plutonium Temperature Effect theoretically described in the literature. Its high level of precision and its “one-step” approach to criticality allowed it to show a significant positive temperature effect for a rather small variation of temperature (+6 °C). The order of magnitude of the effect was confirmed with Monte Carlo calculations using thermal scattering data for hydrogen in the solution produced by IRSN for the purpose of the comparison.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"9 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86741510","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The present study investigated the release of tritium from beryllium irradiated at 323 K to a neutron fluence of 4.67 × 1026 m−2 (E > 1 MeV), corresponding up to 22,000 appm helium and 2000 appm tritium productions. The TPD tests revealed a single tritium release peak during thermal desorption tests, irrespective of the heating mode employed. The tritium release peaks occurred at temperatures ranging from 1031–1136 K, depending on the heating mode, with a desorption energy of 1.6 eV. Additionally, the effective tritium diffusion coefficient was found to vary from 1.2 × 10−12 m2/s at 873 K to 1.8 × 10−10 m2/s at 1073 K. The evolution of beryllium microstructure was found to be dependent on the annealing temperature. No discernible differences were observed between the as-received state and after annealing at 473–773 K for 5 h, with a corresponding porosity range of 1–2%. The annealing at temperatures of 873–1373 K for 5 h resulted in the formation of large bubbles, with porosity increasing sharply above 873 K and reaching 30–60%.
{"title":"Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor","authors":"V. Chakin, R. Rolli, R. Gaisin, W. Van Renterghem","doi":"10.3390/jne4030036","DOIUrl":"https://doi.org/10.3390/jne4030036","url":null,"abstract":"The present study investigated the release of tritium from beryllium irradiated at 323 K to a neutron fluence of 4.67 × 1026 m−2 (E > 1 MeV), corresponding up to 22,000 appm helium and 2000 appm tritium productions. The TPD tests revealed a single tritium release peak during thermal desorption tests, irrespective of the heating mode employed. The tritium release peaks occurred at temperatures ranging from 1031–1136 K, depending on the heating mode, with a desorption energy of 1.6 eV. Additionally, the effective tritium diffusion coefficient was found to vary from 1.2 × 10−12 m2/s at 873 K to 1.8 × 10−10 m2/s at 1073 K. The evolution of beryllium microstructure was found to be dependent on the annealing temperature. No discernible differences were observed between the as-received state and after annealing at 473–773 K for 5 h, with a corresponding porosity range of 1–2%. The annealing at temperatures of 873–1373 K for 5 h resulted in the formation of large bubbles, with porosity increasing sharply above 873 K and reaching 30–60%.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"38 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-08-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79642234","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The heat transfer during pool boiling incorporates a higher rate of heat dissipation capability at low-temperature differences. This technique is widely used in the nuclear industry for thermal management. In this study, the effect of tube diameter and length on critical heat flux (CHF) at atmospheric conditions in saturated water during pool boiling is analyzed. The tubes of SS 304 are kept in the vertical orientation. The diameter of the tubes ranges from 1.2 mm to 9 mm. The tube lengths varying from 50mm to 1000mm. It has been noted that tubes of smaller diameter show a greater magnitude of CHF for the given length. For a given diameter, a longer tube is found to have lower CHF than the ones having lesser length. The variation in the CHF magnitude is negligible for tubes with a diameter of more than 2.5 mm beyond a length of 200 mm. The location of occurrence of CHF is near the bottom end of the vertical tube. The study illustrates the behavior of bubble nucleation for various tube dimensions and heat fluxes. The inception and detachment of bubbles for different tubes are analyzed. The pool boiling regime is categorized and studied basing the behaviour of the incepted and departed bubble. A mathematical relation that empirically accounts for the effect of tube dimensions i.e. length and diameter on pool boiling CHF is proposed. The experimental CHF data obtained during pool boiling are tabulated towards contributing to the CHF databank.
池沸腾过程中的传热在低温差下具有较高的散热能力。该技术广泛应用于核工业的热管理。本文分析了常压条件下饱和水池沸腾过程中管径和管长对临界热流密度的影响。不锈钢管保持在垂直方向。管的直径范围为1.2 mm ~ 9mm。管长从50mm到1000mm不等。已经注意到，对于给定的长度，直径较小的管显示出更大的CHF幅度。对于给定的直径，较长的管子比长度较小的管子具有更低的CHF。对于直径大于2.5 mm且长度超过200 mm的管，CHF大小的变化可以忽略不计。CHF发生的位置在垂直管的下端附近。研究说明了不同管径和不同热流密度下的气泡成核行为。分析了不同管道中气泡的产生和分离。根据入泡和离泡的行为对池沸腾状态进行了分类和研究。本文提出了一种计算管道尺寸(即长度和直径)对池沸腾CHF影响的数学关系。池沸腾过程中获得的实验CHF数据被制成表格，以供CHF数据库使用。
{"title":"Experimental Analysis Of Bubble Behavior And Critical Heat Flux During Pool Boiling On Vertical Circular Tubes","authors":"Bikash Pattanayak, Hardik B. Kothadia","doi":"10.1115/1.4063041","DOIUrl":"https://doi.org/10.1115/1.4063041","url":null,"abstract":"\u0000 The heat transfer during pool boiling incorporates a higher rate of heat dissipation capability at low-temperature differences. This technique is widely used in the nuclear industry for thermal management. In this study, the effect of tube diameter and length on critical heat flux (CHF) at atmospheric conditions in saturated water during pool boiling is analyzed. The tubes of SS 304 are kept in the vertical orientation. The diameter of the tubes ranges from 1.2 mm to 9 mm. The tube lengths varying from 50mm to 1000mm. It has been noted that tubes of smaller diameter show a greater magnitude of CHF for the given length. For a given diameter, a longer tube is found to have lower CHF than the ones having lesser length. The variation in the CHF magnitude is negligible for tubes with a diameter of more than 2.5 mm beyond a length of 200 mm. The location of occurrence of CHF is near the bottom end of the vertical tube. The study illustrates the behavior of bubble nucleation for various tube dimensions and heat fluxes. The inception and detachment of bubbles for different tubes are analyzed. The pool boiling regime is categorized and studied basing the behaviour of the incepted and departed bubble. A mathematical relation that empirically accounts for the effect of tube dimensions i.e. length and diameter on pool boiling CHF is proposed. The experimental CHF data obtained during pool boiling are tabulated towards contributing to the CHF databank.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"31 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87405604","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The steel boxes that contain radioactive solid wastes should be accurately identified and measured. On account of the uneven internal filling and radioactive distribution in the steel boxes that storing radioactive solid wastes, the non-destructive measurement analysis technique can be utilized to perform overall measurement of the wastes, which will not damage the steel boxes and produce secondary wastes. The γ-ray spectrum measurement analysis algorithm of radioactive solid waste steel boxes was set up by way of discrete treatment for γ-ray spectrum measurement analysis algorithm combined with the traditional CT principle in this paper, the progressive split voxel measurement method was applied to conduct discrete treatment for the radioactive solid waste steel boxes, with the influence of voxel interactions on detection efficiency was taken into consideration, the calibration model of detection efficiency was set up through Monte Carlo simulation combined with algebraic reconstruction technique, the inversion correction of efficiency for the physically coincident part of the detected voxel volume was achieved. The proposed algorithm was preliminarily applied on the analyze platform of steel boxes that storing FA-IV- type radioactive solid wastes, and the detection and verification on the standard radioactive source of known activities was also conducted, the findings revealed that the relative deviation between the reconstruction results and truth-value of radioactive activities in the steel boxes is less than 30%, which conformed to the expected results and the proposed algorithm can be popularized and applicated in the detection works in related to other.
{"title":"Study of γ-ray Spectrum Measurement Analysis Algorithm of Radioactive Solid Waste Steel Boxes","authors":"Yurong Li, Lixia He, Jing Wen, Jiewen Shao","doi":"10.1115/1.4063039","DOIUrl":"https://doi.org/10.1115/1.4063039","url":null,"abstract":"\u0000 The steel boxes that contain radioactive solid wastes should be accurately identified and measured. On account of the uneven internal filling and radioactive distribution in the steel boxes that storing radioactive solid wastes, the non-destructive measurement analysis technique can be utilized to perform overall measurement of the wastes, which will not damage the steel boxes and produce secondary wastes. The γ-ray spectrum measurement analysis algorithm of radioactive solid waste steel boxes was set up by way of discrete treatment for γ-ray spectrum measurement analysis algorithm combined with the traditional CT principle in this paper, the progressive split voxel measurement method was applied to conduct discrete treatment for the radioactive solid waste steel boxes, with the influence of voxel interactions on detection efficiency was taken into consideration, the calibration model of detection efficiency was set up through Monte Carlo simulation combined with algebraic reconstruction technique, the inversion correction of efficiency for the physically coincident part of the detected voxel volume was achieved. The proposed algorithm was preliminarily applied on the analyze platform of steel boxes that storing FA-IV- type radioactive solid wastes, and the detection and verification on the standard radioactive source of known activities was also conducted, the findings revealed that the relative deviation between the reconstruction results and truth-value of radioactive activities in the steel boxes is less than 30%, which conformed to the expected results and the proposed algorithm can be popularized and applicated in the detection works in related to other.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"27 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82303388","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Srivastava, Saikrishna Nadella, N. K. Maheshwari
A Molten Salt Natural Circulation Loop (MSNCL) has been setup to study the steady state, transient and stability characteristics of molten nitrate salt, mixture of Sodium Nitrate and Potassium Nitrate in 60:40 ratios by weight. Natural circulation experiments in a temperature range of 290 °C to 600 °C have been performed. A semi-analytical linear model was derived for stability analysis of rectangular natural circulation loops with conventional localized surface heating and cooling. The developed model includes the effect of wall thermal inertia, variable internal heat transfer coefficient, finite secondary side heat transfer coefficient at cooler and heat losses for predicting the stability map. These effects are incorporated considering their significant role in modelling high temperature molten salt based natural circulation systems. The developed model has been first validated with the experimental data of water loop available in literature. The developed model is then validated with the experimental data generated in MSNCL. Validation of in-house developed non-linear model has also been performed against the same experimental data. The comparison of both linear and non-linear stability analysis with the experimental data shows good agreement and articulate the importance of various parameters which have been included in the developed model.
{"title":"Linear And Non-Linear Stability Analysis Of Molten Salt Natural Circulation Loop","authors":"A. Srivastava, Saikrishna Nadella, N. K. Maheshwari","doi":"10.1115/1.4063040","DOIUrl":"https://doi.org/10.1115/1.4063040","url":null,"abstract":"\u0000 A Molten Salt Natural Circulation Loop (MSNCL) has been setup to study the steady state, transient and stability characteristics of molten nitrate salt, mixture of Sodium Nitrate and Potassium Nitrate in 60:40 ratios by weight. Natural circulation experiments in a temperature range of 290 °C to 600 °C have been performed. A semi-analytical linear model was derived for stability analysis of rectangular natural circulation loops with conventional localized surface heating and cooling. The developed model includes the effect of wall thermal inertia, variable internal heat transfer coefficient, finite secondary side heat transfer coefficient at cooler and heat losses for predicting the stability map. These effects are incorporated considering their significant role in modelling high temperature molten salt based natural circulation systems. The developed model has been first validated with the experimental data of water loop available in literature. The developed model is then validated with the experimental data generated in MSNCL. Validation of in-house developed non-linear model has also been performed against the same experimental data. The comparison of both linear and non-linear stability analysis with the experimental data shows good agreement and articulate the importance of various parameters which have been included in the developed model.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"7 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74387250","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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