Abstract The aim of this study is to assess the performance of an existing dry storage cask design for accident tolerant fuel loading case and to examine the compliance with the safety limits applied currently for dry storage. In the study, for a dry storage cask design currently in use, criticality calculations, dose rate evaluation and thermal analyses are performed in case of loading with the accident tolerant fuel discharged from a PWR. Firstly, for an accident tolerant fuel selected among the concepts proposed for use in light water reactors, burnup analyses are performed by utilizing the Serpent continuous energy code and spent fuel characteristics are determined. Then, criticality analyzes are carried out by using the Serpent Monte Carlo code for the case of loading the accident tolerant fuel into the selected dry storage cask design. Gamma and neutron dose rates at the outer surface and close distances of the storage cask are determined with the Serpent code. To evaluate the thermal performance of the storage cask, thermal analyzes are performed by using the ANSYS Fluent computational fluid dynamics code. The analysis results are compared with the nuclear safety criteria applied to dry storage casks. Results of the analysis show that the dry storage cask design currently in-use does not exceed the criticality, dose rate and maximum surface temperature limits when loaded with spent accident tolerant fuel.
{"title":"Performance evaluation of a currently in-use dry storage cask design for spent accident tolerant fuel loading case under normal operation condition","authors":"Habibe Durdu, Banu Bulut Acar","doi":"10.1515/kern-2023-0001","DOIUrl":"https://doi.org/10.1515/kern-2023-0001","url":null,"abstract":"Abstract The aim of this study is to assess the performance of an existing dry storage cask design for accident tolerant fuel loading case and to examine the compliance with the safety limits applied currently for dry storage. In the study, for a dry storage cask design currently in use, criticality calculations, dose rate evaluation and thermal analyses are performed in case of loading with the accident tolerant fuel discharged from a PWR. Firstly, for an accident tolerant fuel selected among the concepts proposed for use in light water reactors, burnup analyses are performed by utilizing the Serpent continuous energy code and spent fuel characteristics are determined. Then, criticality analyzes are carried out by using the Serpent Monte Carlo code for the case of loading the accident tolerant fuel into the selected dry storage cask design. Gamma and neutron dose rates at the outer surface and close distances of the storage cask are determined with the Serpent code. To evaluate the thermal performance of the storage cask, thermal analyzes are performed by using the ANSYS Fluent computational fluid dynamics code. The analysis results are compared with the nuclear safety criteria applied to dry storage casks. Results of the analysis show that the dry storage cask design currently in-use does not exceed the criticality, dose rate and maximum surface temperature limits when loaded with spent accident tolerant fuel.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"2018 1","pages":"424 - 436"},"PeriodicalIF":0.5,"publicationDate":"2023-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73335610","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Prateek D. Malwe, Aarti Mukayanamath, H. Panchal, N. Gupta, C. Prakash, M. M. Abdul Zahra
Abstract Heat transfer enhancement is required for numerous situations, i.e., gas turbines, nuclear power plants, micro and macro scale heat transfer, airfoil cooling, electronic cooling, semiconductors, biomedical and combustion chamber lines, etc. One of the prominent ways of increasing the heat transfer coefficient from the surface of a heat exchanger is by moving the position of the thermal boundary layer to make it either thinner or break the same partially. It requires making use of an increased surface area/fins. Accordingly, the research progressed in heat transfer enhancement by using concavities/dimples of the heat exchanger surfaces to improve the heat transfer coefficient and heat transfer rate. These impregnations are made on the internal flow tubes/surfaces of the heat exchanger surfaces. The present research work aims at the experimental investigation of a heat exchanger to determine the airflow pattern and computation of heat transfer rate on the dimpled surfaces. This research work will be beneficial and applicable to heat transfer enhancement applications like micro heat transfer, where space constraint is considered. The geometries considered for the experiment include flat plates and dimpled surfaces. The parameters like Reynolds number (varied from 20,000 to 50,000), dimple depth to imprint diameter ratio (varied from 0.2 to 0.4), and heater input to the test plates (varied from 75 to 120 W) are considered for the comparisons. The results with dimpled surfaces are compared with the flat plate surfaces having no dimples. The Reynolds and Nusselt numbers rise in direct proportion to the heater input. For pin fin and dimpled plate, the ratio of Nusselt number to area average Nusselt number drops for 75 W and 100 W input. The dimpled plate with a ratio of 0.3 between imprint diameter to dimple depth had the highest ratio of Nusselt number to Nusselt number value for the entire group.
{"title":"Heat transfer enhancement of heat exchanger using rectangular channel with cavities","authors":"Prateek D. Malwe, Aarti Mukayanamath, H. Panchal, N. Gupta, C. Prakash, M. M. Abdul Zahra","doi":"10.1515/kern-2023-0032","DOIUrl":"https://doi.org/10.1515/kern-2023-0032","url":null,"abstract":"Abstract Heat transfer enhancement is required for numerous situations, i.e., gas turbines, nuclear power plants, micro and macro scale heat transfer, airfoil cooling, electronic cooling, semiconductors, biomedical and combustion chamber lines, etc. One of the prominent ways of increasing the heat transfer coefficient from the surface of a heat exchanger is by moving the position of the thermal boundary layer to make it either thinner or break the same partially. It requires making use of an increased surface area/fins. Accordingly, the research progressed in heat transfer enhancement by using concavities/dimples of the heat exchanger surfaces to improve the heat transfer coefficient and heat transfer rate. These impregnations are made on the internal flow tubes/surfaces of the heat exchanger surfaces. The present research work aims at the experimental investigation of a heat exchanger to determine the airflow pattern and computation of heat transfer rate on the dimpled surfaces. This research work will be beneficial and applicable to heat transfer enhancement applications like micro heat transfer, where space constraint is considered. The geometries considered for the experiment include flat plates and dimpled surfaces. The parameters like Reynolds number (varied from 20,000 to 50,000), dimple depth to imprint diameter ratio (varied from 0.2 to 0.4), and heater input to the test plates (varied from 75 to 120 W) are considered for the comparisons. The results with dimpled surfaces are compared with the flat plate surfaces having no dimples. The Reynolds and Nusselt numbers rise in direct proportion to the heater input. For pin fin and dimpled plate, the ratio of Nusselt number to area average Nusselt number drops for 75 W and 100 W input. The dimpled plate with a ratio of 0.3 between imprint diameter to dimple depth had the highest ratio of Nusselt number to Nusselt number value for the entire group.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"1 1","pages":"532 - 540"},"PeriodicalIF":0.5,"publicationDate":"2023-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91316437","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Rui Yu, Guan Wang, Wei Jiang, C. Yao, Lu Zhang, L. Gu
Abstract The radial non-uniformity of an ADS fuel rod power density on the peak plane is as high as 10.46 %, which is derived from the reactor physics calculations. In order to investigate the influence of the above-mentioned radial power inhomogeneity on the fuel temperature distribution, this paper constructs a set of two-dimensional comparison examples, where one example uses a uniform heat source and another uses a non-uniform heat source distribution, with taking the fuel segment and its corresponding cladding segment in the region where the peak plane is located as the research object. The finite element software COMSOL is used to conduct the heat transfer analysis. The results of the study show that the fuel temperature under radial non-uniform power distribution is almost the same as that under uniform power. Therefore, this radial non-uniformity can be completely ignored when the research object of temperature is considered. The quantitative calculation carried out in this research can provide certain data support for the engineering research of accelerator driven subcritical system, and can also provide certain guidance for the performance analysis of such fuel elements.
{"title":"Evaluating the influence of radial power heterogeneity of fuel rod on its temperature in an accelerator driven subcritical system","authors":"Rui Yu, Guan Wang, Wei Jiang, C. Yao, Lu Zhang, L. Gu","doi":"10.1515/kern-2023-0020","DOIUrl":"https://doi.org/10.1515/kern-2023-0020","url":null,"abstract":"Abstract The radial non-uniformity of an ADS fuel rod power density on the peak plane is as high as 10.46 %, which is derived from the reactor physics calculations. In order to investigate the influence of the above-mentioned radial power inhomogeneity on the fuel temperature distribution, this paper constructs a set of two-dimensional comparison examples, where one example uses a uniform heat source and another uses a non-uniform heat source distribution, with taking the fuel segment and its corresponding cladding segment in the region where the peak plane is located as the research object. The finite element software COMSOL is used to conduct the heat transfer analysis. The results of the study show that the fuel temperature under radial non-uniform power distribution is almost the same as that under uniform power. Therefore, this radial non-uniformity can be completely ignored when the research object of temperature is considered. The quantitative calculation carried out in this research can provide certain data support for the engineering research of accelerator driven subcritical system, and can also provide certain guidance for the performance analysis of such fuel elements.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"99 1","pages":"527 - 531"},"PeriodicalIF":0.5,"publicationDate":"2023-07-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83336452","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract This investigation reports on the experimental outcomes of the pool boiling heat transfer characteristics, specifically on the downward heated surface, concerning reverse osmosis water and γ-Fe2O3 nanofluids. To conduct the pool boiling experiments, γ-Fe2O3 nanofluids were prepared with variable concentrations ranging from 2 mg/L to 10 mg/L. Analysis of the experimental data revealed that a concentration of 5 mg/L yielded the greatest enhancement effect on critical heat flux (CHF), with an increase of 13.5 %. However, the results also indicated that excessively high concentrations of nanofluid had a negative impact on CHF enhancement. The impact of nanofluids on heat transfer performance was investigated by analyzing the observed bubble behavior during the boiling process, measuring the drop angle and surface roughness post-experiment, and characterizing the heated surface morphology via scanning electron microscopy (SEM). Through these methods, the underlying mechanism behind the impact of nanofluids on heat transfer performance was identified and analyzed.
{"title":"Experimental study on boiling heat transfer of γ-Fe2O3 nanofluids on a downward heated surface","authors":"Jia Gao, H. Hsieh, Songling Liu, Xintian Cai, Sai-lan Wang, Shiqi Wang, Shihao Zhang, Zhusheng Guo","doi":"10.1515/kern-2023-0033","DOIUrl":"https://doi.org/10.1515/kern-2023-0033","url":null,"abstract":"Abstract This investigation reports on the experimental outcomes of the pool boiling heat transfer characteristics, specifically on the downward heated surface, concerning reverse osmosis water and γ-Fe2O3 nanofluids. To conduct the pool boiling experiments, γ-Fe2O3 nanofluids were prepared with variable concentrations ranging from 2 mg/L to 10 mg/L. Analysis of the experimental data revealed that a concentration of 5 mg/L yielded the greatest enhancement effect on critical heat flux (CHF), with an increase of 13.5 %. However, the results also indicated that excessively high concentrations of nanofluid had a negative impact on CHF enhancement. The impact of nanofluids on heat transfer performance was investigated by analyzing the observed bubble behavior during the boiling process, measuring the drop angle and surface roughness post-experiment, and characterizing the heated surface morphology via scanning electron microscopy (SEM). Through these methods, the underlying mechanism behind the impact of nanofluids on heat transfer performance was identified and analyzed.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"148 1","pages":"518 - 526"},"PeriodicalIF":0.5,"publicationDate":"2023-06-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88663456","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
S. Sriyono, D. Saprudin, M. Rafi, G. R. Sunaryo, Nugraha Luhur, F. A. Muslimu
Abstract The liquid radioactive waste generated by the G.A. Siwabessy reactor (RSG-GAS) is categorized into low-activity liquid radwaste (LALR) and medium-activity liquid radwaste (MALR). The radionuclide content of both LALR and MALR can use as an indicator of the structural integrity of the reactor’s systems, structures, and components (SSC). To evaluate the degradation of the reactor SSC, the radionuclide species were identified, and their activities were measured using gamma spectroscopy. Based on the identified radionuclides, the process of their formation can be traced. The radionuclides identified in LALR were 24Na, 51Cr, 59Fe, 60Co, 65Zn, and 124Sb, while the radionuclides in MALR were 24Na, 51Cr, 58Co, 59Fe, 60Co, 65Ni, 65Zn, 89Kr, 90Kr, 109Cd, 131I, 132I, 140Ba, 137Cs, 146Ce, and several others. The radionuclides found can be classified into corrosion product activation (60Co, 65Zn, 51Cr, 59Fe, 24Na, 65Ni), topaz impurities activation (51Cr, 59Fe, 60Co, 65Zn), fission product (90Kr, 140Ba, 131I, 137Cs, etc.), and demineralized water impurities activation (51Cr, 59Fe, 65Zn, 60Co, etc.). After comparing the activity value of each radionuclide with the limit value in the safety analysis report document, we can conclude that the activity of each one is below the required level. It can infer that the structural integrity of reactor SSC is still well maintain. During routine monitoring, the radionuclide content in the primary coolant fluctuates depending on the reactor load. The concentration of radionuclides detected varies when a large or small number of research samples are loaded onto the core. Nevertheless, their activities remain within the required safety limits.
{"title":"Identification and tracing of radionuclides in low- and medium-activity liquid radwaste sources of G.A. Siwabessy reactor","authors":"S. Sriyono, D. Saprudin, M. Rafi, G. R. Sunaryo, Nugraha Luhur, F. A. Muslimu","doi":"10.1515/kern-2022-0113","DOIUrl":"https://doi.org/10.1515/kern-2022-0113","url":null,"abstract":"Abstract The liquid radioactive waste generated by the G.A. Siwabessy reactor (RSG-GAS) is categorized into low-activity liquid radwaste (LALR) and medium-activity liquid radwaste (MALR). The radionuclide content of both LALR and MALR can use as an indicator of the structural integrity of the reactor’s systems, structures, and components (SSC). To evaluate the degradation of the reactor SSC, the radionuclide species were identified, and their activities were measured using gamma spectroscopy. Based on the identified radionuclides, the process of their formation can be traced. The radionuclides identified in LALR were 24Na, 51Cr, 59Fe, 60Co, 65Zn, and 124Sb, while the radionuclides in MALR were 24Na, 51Cr, 58Co, 59Fe, 60Co, 65Ni, 65Zn, 89Kr, 90Kr, 109Cd, 131I, 132I, 140Ba, 137Cs, 146Ce, and several others. The radionuclides found can be classified into corrosion product activation (60Co, 65Zn, 51Cr, 59Fe, 24Na, 65Ni), topaz impurities activation (51Cr, 59Fe, 60Co, 65Zn), fission product (90Kr, 140Ba, 131I, 137Cs, etc.), and demineralized water impurities activation (51Cr, 59Fe, 65Zn, 60Co, etc.). After comparing the activity value of each radionuclide with the limit value in the safety analysis report document, we can conclude that the activity of each one is below the required level. It can infer that the structural integrity of reactor SSC is still well maintain. During routine monitoring, the radionuclide content in the primary coolant fluctuates depending on the reactor load. The concentration of radionuclides detected varies when a large or small number of research samples are loaded onto the core. Nevertheless, their activities remain within the required safety limits.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"156 1","pages":"413 - 423"},"PeriodicalIF":0.5,"publicationDate":"2023-06-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82910382","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract The sodium-cooled fast reactor is a Generation-IV International Forum recommended technology, with an aim to improve sustainability, safety, and proliferation resistance. To ensure accurate reactor physics calculation and safety analyses, nuclear data libraries require continuous improvement through modifications based on additional measurements, evaluations, and validation studies with criticality experiments. In this work the Sodium-cooled Fast Reactor Uncertainty Analysis in Modeling (SFR-UAM) benchmark served as a basis to assess differences in nuclear data libraries and estimate variability in criticality and power distribution results. The research has been carried out using the OpenMC code and the study presented here covers two SFR models: MOX-3600 and ABR-1000. The neutronic calculation of numerous parameters in fast spectrum systems including effective multiplication factor (keff), effective delayed neutron fraction (βeff), sodium void reactivity (ΔρNa), Doppler constant (ΔρDoppler), and control rod (ρCR) worth were calculated and compared mainly to five libraries: ENDF/B-VII.1, ENDF/B-VIII, JEFF-3.3, JENDL-4.0 and TENDL-2019. In addition, sensitivity calculations using GPT-free method were conducted to understand relevant sensitivities for a given quantity of interest in major isotope/reaction pairs. The major driver of observed uncertainty in keff are found for the high actinide isotopes mainly capture cross section of 239, 240Pu as well as fission reaction of 239Pu.
{"title":"Neutronic analysis of the European sodium cooled fast reactor with Monte Carlo code OpenMC","authors":"Md. Ariful Islam","doi":"10.1515/kern-2023-0016","DOIUrl":"https://doi.org/10.1515/kern-2023-0016","url":null,"abstract":"Abstract The sodium-cooled fast reactor is a Generation-IV International Forum recommended technology, with an aim to improve sustainability, safety, and proliferation resistance. To ensure accurate reactor physics calculation and safety analyses, nuclear data libraries require continuous improvement through modifications based on additional measurements, evaluations, and validation studies with criticality experiments. In this work the Sodium-cooled Fast Reactor Uncertainty Analysis in Modeling (SFR-UAM) benchmark served as a basis to assess differences in nuclear data libraries and estimate variability in criticality and power distribution results. The research has been carried out using the OpenMC code and the study presented here covers two SFR models: MOX-3600 and ABR-1000. The neutronic calculation of numerous parameters in fast spectrum systems including effective multiplication factor (keff), effective delayed neutron fraction (βeff), sodium void reactivity (ΔρNa), Doppler constant (ΔρDoppler), and control rod (ρCR) worth were calculated and compared mainly to five libraries: ENDF/B-VII.1, ENDF/B-VIII, JEFF-3.3, JENDL-4.0 and TENDL-2019. In addition, sensitivity calculations using GPT-free method were conducted to understand relevant sensitivities for a given quantity of interest in major isotope/reaction pairs. The major driver of observed uncertainty in keff are found for the high actinide isotopes mainly capture cross section of 239, 240Pu as well as fission reaction of 239Pu.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"36 1","pages":"399 - 412"},"PeriodicalIF":0.5,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72681280","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract In case of a postulated severe accident in a water-cooled nuclear power plant significant amounts of hydrogen (H2) and carbon monoxide (CO) can be generated and released into the containment or reactor building where it might form a combustible mixture with air assuming passive autocatalytic recombiners are not available. In case of ignition, pressure peaks might occur, that are relevant for the integrity of safety relevant equipment and the containment or reactor building. It is therefore important for safety analysis to be able to correctly predict combustion phenomena that might occur. The accident analysis code AC2 2021.0 which is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) includes the Containment Code System (COCOSYS version 3.1) for the simulation of containment phenomena. COCOSYS contains the model FRONT for the simulation of premixed deflagration of H2 and CO. Recent code validation using H2 deflagration tests conducted in the multi-compartment THAI+ test facility shows that the flame propagation stops prematurely in simulations of some tests. This is partly attributed to the missing separation of burned and unburned atmosphere which leads to a reduction in fuel concentration in not yet burning zones connected to a burning zone. Model improvement potential was identified which is addressed in this paper. A model extension to separate burned and unburned atmosphere via a junction model is proposed and implemented into a development version of COCOSYS 3.1. First validation results using the THAI test HD-39 are discussed that show improved prediction capability by the extended model.
摘要在水冷核电站发生严重事故的情况下，大量的氢(H2)和一氧化碳(CO)会产生并释放到安全壳或反应堆建筑中，如果没有被动的自催化重组器，它们可能与空气形成可燃混合物。在点火的情况下，可能会出现压力峰值，这与安全相关设备和安全壳或反应堆建筑物的完整性有关。因此，能够正确预测可能发生的燃烧现象对于安全分析非常重要。事故分析代码AC2 2021.0是由Gesellschaft fr Anlagen- und Reaktorsicherheit (GRS)开发的，其中包括用于模拟安全壳现象的安全壳代码系统(COCOSYS版本3.1)。COCOSYS包含用于模拟H2和CO预混爆燃的FRONT模型。最近在多室THAI+测试设施中进行的H2爆燃测试验证代码表明，在模拟某些测试中火焰传播提前停止。这在一定程度上是由于未将燃烧的大气和未燃烧的大气分离，导致与燃烧区相连的尚未燃烧区的燃料浓度降低。确定了模型改进的潜力，并在本文中进行了讨论。提出了一种模型扩展，通过结点模型来分离燃烧和未燃烧的大气，并在COCOSYS 3.1的开发版本中实现。讨论了使用THAI测试HD-39的首次验证结果，表明扩展模型提高了预测能力。
{"title":"An approach for an extension of the deflagration model in containment code system COCOSYS to separate burned and unburned atmosphere via junctions","authors":"Johannes Hoffrichter, M. Koch","doi":"10.1515/kern-2023-0021","DOIUrl":"https://doi.org/10.1515/kern-2023-0021","url":null,"abstract":"Abstract In case of a postulated severe accident in a water-cooled nuclear power plant significant amounts of hydrogen (H2) and carbon monoxide (CO) can be generated and released into the containment or reactor building where it might form a combustible mixture with air assuming passive autocatalytic recombiners are not available. In case of ignition, pressure peaks might occur, that are relevant for the integrity of safety relevant equipment and the containment or reactor building. It is therefore important for safety analysis to be able to correctly predict combustion phenomena that might occur. The accident analysis code AC2 2021.0 which is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) includes the Containment Code System (COCOSYS version 3.1) for the simulation of containment phenomena. COCOSYS contains the model FRONT for the simulation of premixed deflagration of H2 and CO. Recent code validation using H2 deflagration tests conducted in the multi-compartment THAI+ test facility shows that the flame propagation stops prematurely in simulations of some tests. This is partly attributed to the missing separation of burned and unburned atmosphere which leads to a reduction in fuel concentration in not yet burning zones connected to a burning zone. Model improvement potential was identified which is addressed in this paper. A model extension to separate burned and unburned atmosphere via a junction model is proposed and implemented into a development version of COCOSYS 3.1. First validation results using the THAI test HD-39 are discussed that show improved prediction capability by the extended model.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"80 1","pages":"385 - 398"},"PeriodicalIF":0.5,"publicationDate":"2023-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82272596","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Ayesha Alam, Shahab Ud-Din Khan, Muhammad Abdullah, R. Khan, Muhammad Ilyas, Khurram Saleem Chaudri, Ahmad Ali, Sehrish Shakir, Z. Rehman, Shahzaib Zahid, R. Ali
Abstract Handling the power deposition, reducing erosion effects, and plasma configuration are the key factors in the design of a divertor. The design of Pakistan Spherical Tokamak (PST) is based on double-null divertor configuration with actively cooled graphite targets at outer/inner strike point and peak heat flux range capacity of 0.1–0.3 MW/m2. The configuration of PST divertor module is designed with mock-up (used flat type tiles on baffles and dome) and cassette (support PFC and cooling channels) technology. Helium-cooled stage and water-cooled stage are two options for divertor. Therefore, one part of this research is focused on water-cooling system for the divertor. This paper presents the divertor design for PST with cooling channel and material analysis of the divertor, which is carried out in three phases. In the first phase, the plasma edge temperature, density, particle velocity, input power, heat flux, and surface temperature are estimated. In second phase, physics and engineering design of divertor system has been performed. In the third phase, COMSOL simulation has been performed to analyses the material properties, surface temperature rise (∆T °C) at stable heat flux, and thermal hydraulic system for the divertor. It is found from the analysis that the specific heat flux of 0.3 MW/m2 up to 3 s is the safe zone limit. The R & D work ratifies that manufacturing and installation processes are plausible for the proposed divertor design. This design is able to meet the requirement of PST. However, increasing time or specific heat flux beyond these limits would require redesigning of the cooling channel.
{"title":"Optimization of divertor design for Pakistan spherical tokamak","authors":"Ayesha Alam, Shahab Ud-Din Khan, Muhammad Abdullah, R. Khan, Muhammad Ilyas, Khurram Saleem Chaudri, Ahmad Ali, Sehrish Shakir, Z. Rehman, Shahzaib Zahid, R. Ali","doi":"10.1515/kern-2022-0105","DOIUrl":"https://doi.org/10.1515/kern-2022-0105","url":null,"abstract":"Abstract Handling the power deposition, reducing erosion effects, and plasma configuration are the key factors in the design of a divertor. The design of Pakistan Spherical Tokamak (PST) is based on double-null divertor configuration with actively cooled graphite targets at outer/inner strike point and peak heat flux range capacity of 0.1–0.3 MW/m2. The configuration of PST divertor module is designed with mock-up (used flat type tiles on baffles and dome) and cassette (support PFC and cooling channels) technology. Helium-cooled stage and water-cooled stage are two options for divertor. Therefore, one part of this research is focused on water-cooling system for the divertor. This paper presents the divertor design for PST with cooling channel and material analysis of the divertor, which is carried out in three phases. In the first phase, the plasma edge temperature, density, particle velocity, input power, heat flux, and surface temperature are estimated. In second phase, physics and engineering design of divertor system has been performed. In the third phase, COMSOL simulation has been performed to analyses the material properties, surface temperature rise (∆T °C) at stable heat flux, and thermal hydraulic system for the divertor. It is found from the analysis that the specific heat flux of 0.3 MW/m2 up to 3 s is the safe zone limit. The R & D work ratifies that manufacturing and installation processes are plausible for the proposed divertor design. This design is able to meet the requirement of PST. However, increasing time or specific heat flux beyond these limits would require redesigning of the cooling channel.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"82 1","pages":"437 - 445"},"PeriodicalIF":0.5,"publicationDate":"2023-06-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89534053","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract In line with population expansion and industrial development, the world’s energy consumption has been rising gradually over the past three decades. As a result, methods for energy conservation have been sought. One of the most common strategies is heat recovery, which is efficient and cost-effective to the extent possible. Heat recovery is not just about saving energy for primary consumption; it is also about lowering emissions and protecting the environment. In this respect, one of the most important strategies for heat recovery is to develop heat exchangers and exploit the energy associated with many of the processes’ output products in order to use it in new processes. Many researchers working in the field of heat engineering are now looking into novel heat transfer techniques. Use of the heat exchanger as a compact is one of these ways that might be considered. The current review therefore concentrates on the design of plate-fin heat exchangers (PFHE) and multi-stream plate-fin heat exchangers (MSPFHE) based on various models. The current review offers some suggestions for upcoming studies on improving heat transfer and minimizing power use.
{"title":"An investigation of multistream plate-fin heat exchanger modelling and design: a review","authors":"N. O. M. Alyaseen, Salem Mehrzad, M. Saffarian","doi":"10.1515/kern-2022-0119","DOIUrl":"https://doi.org/10.1515/kern-2022-0119","url":null,"abstract":"Abstract In line with population expansion and industrial development, the world’s energy consumption has been rising gradually over the past three decades. As a result, methods for energy conservation have been sought. One of the most common strategies is heat recovery, which is efficient and cost-effective to the extent possible. Heat recovery is not just about saving energy for primary consumption; it is also about lowering emissions and protecting the environment. In this respect, one of the most important strategies for heat recovery is to develop heat exchangers and exploit the energy associated with many of the processes’ output products in order to use it in new processes. Many researchers working in the field of heat engineering are now looking into novel heat transfer techniques. Use of the heat exchanger as a compact is one of these ways that might be considered. The current review therefore concentrates on the design of plate-fin heat exchangers (PFHE) and multi-stream plate-fin heat exchangers (MSPFHE) based on various models. The current review offers some suggestions for upcoming studies on improving heat transfer and minimizing power use.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"19 1","pages":"457 - 474"},"PeriodicalIF":0.5,"publicationDate":"2023-06-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84928550","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"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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