An analysis of the peculiarities of the German Uranium Project (1939–1945) reveals that it was, in many ways, different from what one would expect. There was no work at all on a possible bomb, nor on plutonium. The reactor experiments were limited to subcritical systems and did not attempt to achieve the proclaimed goal of a self-sustaining chain reaction. The so-far identified deficits (lack of interest in Nazi circles, mismanagement, scientific mistakes, and deteriorating work conditions during the war) are relevant but not sufficient for explaining the peculiarities. We deduce that the scientists involved, and even the Heereswaffenamt (army ordnance), shied away from making progress, not only towards a bomb but even towards a reactor. They did not fail; they rather renounced a possible success in order not to provoke political interest in the development of a bomb.
{"title":"The Peculiarities of the German Uranium Project (1939–1945)","authors":"Manfred Popp, Piet de Klerk","doi":"10.3390/jne4030040","DOIUrl":"https://doi.org/10.3390/jne4030040","url":null,"abstract":"An analysis of the peculiarities of the German Uranium Project (1939–1945) reveals that it was, in many ways, different from what one would expect. There was no work at all on a possible bomb, nor on plutonium. The reactor experiments were limited to subcritical systems and did not attempt to achieve the proclaimed goal of a self-sustaining chain reaction. The so-far identified deficits (lack of interest in Nazi circles, mismanagement, scientific mistakes, and deteriorating work conditions during the war) are relevant but not sufficient for explaining the peculiarities. We deduce that the scientists involved, and even the Heereswaffenamt (army ordnance), shied away from making progress, not only towards a bomb but even towards a reactor. They did not fail; they rather renounced a possible success in order not to provoke political interest in the development of a bomb.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-09-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135739742","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}
Rhenium-186 (Re-186) has attracted attention as a medical isotope. The feasibility of producing Re-185, the raw material for Re-186, using a fast reactor was evaluated using a continuous energy Monte Carlo code. The irradiation of natural tantalum (Ta) in the fast reactor can produce Re-185 with an isotopic purity of 99%. A two-step irradiation process with different moderators was found to improve the production rate of Re-185. Specifically, this can be achieved by using zirconium hydride (ZrH1.7) as a moderator in the first transmutation process from natural Ta to tungsten (W), and then zirconium deuteride (ZrD1.7) as a moderator in the second transmutation process from W to Re-185. Due to the two-step irradiation, the production rate of Re-185 from Ta can be increased up to a maximum of 470 times compared with irradiation without a moderator, and 2.3 g of Re-185 can be obtained from 1571 g of Ta in 1 year of irradiation. The proposed isotope production method is a new method that is different from the conventional electromagnetic enrichment process.
铼-186 (Re-186)作为一种医用同位素备受关注。利用连续能量蒙特卡罗程序对快中子反应堆生产Re-186原料Re-185的可行性进行了评估。在快堆中辐照天然钽(Ta)可产生同位素纯度为99%的Re-185。采用两步辐照工艺，采用不同的慢化剂可提高Re-185的收率。具体来说，这可以通过氢化锆(ZrH1.7)作为从天然Ta到钨(W)的第一次嬗变过程的慢化剂，然后氘化锆(ZrD1.7)作为从W到Re-185的第二次嬗变过程的慢化剂来实现。由于采用了两步辐照，与不加慢化剂的辐照相比，Ta产生Re-185的速率最多可提高470倍，在1年的辐照中，1571 g Ta可产生2.3 g Re-185。提出的同位素生产方法是一种不同于传统电磁富集方法的新方法。
{"title":"Feasibility Study on Production of High-Purity Rhenium-185 by Nuclear Transmutation of Natural Tantalum","authors":"Yuki Tanoue, Tsugio Yokoyama, Masaki Ozawa","doi":"10.3390/jne4030039","DOIUrl":"https://doi.org/10.3390/jne4030039","url":null,"abstract":"Rhenium-186 (Re-186) has attracted attention as a medical isotope. The feasibility of producing Re-185, the raw material for Re-186, using a fast reactor was evaluated using a continuous energy Monte Carlo code. The irradiation of natural tantalum (Ta) in the fast reactor can produce Re-185 with an isotopic purity of 99%. A two-step irradiation process with different moderators was found to improve the production rate of Re-185. Specifically, this can be achieved by using zirconium hydride (ZrH1.7) as a moderator in the first transmutation process from natural Ta to tungsten (W), and then zirconium deuteride (ZrD1.7) as a moderator in the second transmutation process from W to Re-185. Due to the two-step irradiation, the production rate of Re-185 from Ta can be increased up to a maximum of 470 times compared with irradiation without a moderator, and 2.3 g of Re-185 can be obtained from 1571 g of Ta in 1 year of irradiation. The proposed isotope production method is a new method that is different from the conventional electromagnetic enrichment process.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82270107","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}
Optimization of dynamical multibeam neutron cancer therapy has recently been shown to be possible via employment of the beam frequencies of neutron waves as a control variable. The concepts of transfer functions, addressed in this paper, can be essential ingredients of such optimization. Accordingly, we study the dynamics of a 1D monochromatic neutron density wave generated by time modulation of a boundary neutron current. It is demonstrated that a certain temporal transfer function of both parabolic (diffusion) and low frequency hyperbolic (P?1 transport) interfacial neutron density wave happens to be frequency non-invariant with a vibrating boundary neutron current. It is proved that, only at high frequencies, both parabolic and hyperbolic interfacial neutron waves turn out to have a fully frequency-invariant and time-invariant temporal transfer function relative to such a vibrating neutron beam at the boundary. The frequency response of an associated complex transfer function is studied and demonstrated to change behavior, from a lag compensator to a fixed gain amplifier, with changing the frequency, neutron absorption and employed theory for neutron diffusion. A highlight of this paper is its illustration that mere continuity of these transfer functions can be a reflection of the correctness of the transport theory employed for modeling the neutron density waves.
{"title":"On Frequency / Time Invariance of Certain Temporal and Complex Transfer Functions for the 1d Interfacial Monochromatic Neutron Density Wave","authors":"N. Haidar","doi":"10.1115/1.4063291","DOIUrl":"https://doi.org/10.1115/1.4063291","url":null,"abstract":"\u0000 Optimization of dynamical multibeam neutron cancer therapy has recently been shown to be possible via employment of the beam frequencies of neutron waves as a control variable. The concepts of transfer functions, addressed in this paper, can be essential ingredients of such optimization. Accordingly, we study the dynamics of a 1D monochromatic neutron density wave generated by time modulation of a boundary neutron current. It is demonstrated that a certain temporal transfer function of both parabolic (diffusion) and low frequency hyperbolic (P?1 transport) interfacial neutron density wave happens to be frequency non-invariant with a vibrating boundary neutron current. It is proved that, only at high frequencies, both parabolic and hyperbolic interfacial neutron waves turn out to have a fully frequency-invariant and time-invariant temporal transfer function relative to such a vibrating neutron beam at the boundary. The frequency response of an associated complex transfer function is studied and demonstrated to change behavior, from a lag compensator to a fixed gain amplifier, with changing the frequency, neutron absorption and employed theory for neutron diffusion. A highlight of this paper is its illustration that mere continuity of these transfer functions can be a reflection of the correctness of the transport theory employed for modeling the neutron density waves.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79158023","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}
This paper describes the development, analysis, testing of a RELAP5-3D system thermal-hydraulics model for a 50-MWel-class pressurized water reactor - small modular reactor (PWR-SMR), similar to that by NuScale Power. This study focuses on a series of sensitivity tests to investigate the impacts of model changes. Parameters considered in the sensitivity study included the surge line junction resistance (SLJR), steam generator (SG) heat transfer area (SGHTA), SG primary flow area (SGPFA), SG secondary pressure (SGSP), and SG secondary flow rate (SGSF). Results for the reference and sensitivity simulations are compared with available design data. The flow in the primary circuit of the PWR-SMR is driven by natural circulation, and can be sensitive to changes in hydraulic resistance and pressure drop in system components. Initial results demonstrated significant flow oscillations. As a result of sensitivity studies, it was found that the surge line junction resistance needed to be increased to a factor of 30 to reduce mass flow oscillations to less than ±2%. Modifications to the steam generator heat transfer area, primary flow area, or secondary pressure have very little impact in reducing flow oscillations. However, it was found that the steam generator secondary flow rate will affect primary circuit flow oscillations, and when the SGSF was artificially increased from 68 kg/s (design data) to 91 kg/s (a 36% increase), the oscillations were eliminated, along with better matching with design data for core flow rate and inlet/outlet temperatures.
{"title":"“Development And Testing Of A System Thermal-Hydraulics Model For A 50-Mwel-Class Pressurized Water Reactor - Small Modular Reactor (Pwr-Smr)”","authors":"Shu Jun Wang, Xianmin Huang, Y. Rao, B. Bromley","doi":"10.1115/1.4063240","DOIUrl":"https://doi.org/10.1115/1.4063240","url":null,"abstract":"\u0000 This paper describes the development, analysis, testing of a RELAP5-3D system thermal-hydraulics model for a 50-MWel-class pressurized water reactor - small modular reactor (PWR-SMR), similar to that by NuScale Power. This study focuses on a series of sensitivity tests to investigate the impacts of model changes. Parameters considered in the sensitivity study included the surge line junction resistance (SLJR), steam generator (SG) heat transfer area (SGHTA), SG primary flow area (SGPFA), SG secondary pressure (SGSP), and SG secondary flow rate (SGSF). Results for the reference and sensitivity simulations are compared with available design data. The flow in the primary circuit of the PWR-SMR is driven by natural circulation, and can be sensitive to changes in hydraulic resistance and pressure drop in system components. Initial results demonstrated significant flow oscillations. As a result of sensitivity studies, it was found that the surge line junction resistance needed to be increased to a factor of 30 to reduce mass flow oscillations to less than ±2%. Modifications to the steam generator heat transfer area, primary flow area, or secondary pressure have very little impact in reducing flow oscillations. However, it was found that the steam generator secondary flow rate will affect primary circuit flow oscillations, and when the SGSF was artificially increased from 68 kg/s (design data) to 91 kg/s (a 36% increase), the oscillations were eliminated, along with better matching with design data for core flow rate and inlet/outlet temperatures.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-08-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88457333","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}
B. Gera, P. Sharma, A. Karanam, Shubham Mishra, T. I., Nandan Saha, A. Dutta, P. Goyal, V. Verma, J. Chattopadhyay
A significant amount of hydrogen may be released inside the containment of water cooled nuclear power reactor under postulated accident conditions. Its distribution in the multi-compartment containment geometry must be known to manage and mitigate the local hydrogen concentration in combustible pockets. An experimental study to characterize the behavior of a lighter gas (helium in place of hydrogen) in a multi-compartment Containment Studies Facility (CSF) has been pursued. Helium distribution experiments have been performed in CSF by varying important accident parameters like helium release rate, duration and injection area. The experimental studies performed in CSF depict helium stratification in the upper dome region. Stratification in terms of stratification/effective stratification factor determined for a range of experiments. The present experimental studies are important for understanding hydrogen distribution characteristics in multi-compartment containment geometry and benchmarking of CFD codes. Based on these studies some important prevailing practices for recombiner placement were endorsed.
{"title":"Experimental Study Of Helium Stratification In Multi-Compartment Containment Studies Facility","authors":"B. Gera, P. Sharma, A. Karanam, Shubham Mishra, T. I., Nandan Saha, A. Dutta, P. Goyal, V. Verma, J. Chattopadhyay","doi":"10.1115/1.4063239","DOIUrl":"https://doi.org/10.1115/1.4063239","url":null,"abstract":"\u0000 A significant amount of hydrogen may be released inside the containment of water cooled nuclear power reactor under postulated accident conditions. Its distribution in the multi-compartment containment geometry must be known to manage and mitigate the local hydrogen concentration in combustible pockets. An experimental study to characterize the behavior of a lighter gas (helium in place of hydrogen) in a multi-compartment Containment Studies Facility (CSF) has been pursued. Helium distribution experiments have been performed in CSF by varying important accident parameters like helium release rate, duration and injection area. The experimental studies performed in CSF depict helium stratification in the upper dome region. Stratification in terms of stratification/effective stratification factor determined for a range of experiments. The present experimental studies are important for understanding hydrogen distribution characteristics in multi-compartment containment geometry and benchmarking of CFD codes. Based on these studies some important prevailing practices for recombiner placement were endorsed.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-08-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91273514","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}
Many advanced reactor concept designs rely on high-assay low-enriched uranium (HALEU) fuel, enriched up to approximately 19.75% 235U by weight. Efforts are underway by the US government to increase HALEU production in the United States to meet anticipated needs. However, very few data exist for validation of computational models that include HALEU, beyond a few fresh fuel benchmark specifications in the International Reactor Physics Experiment Evaluation Project. Nevertheless, there are other data with potential value available for developing into quality benchmarks for use in data- and software-validation efforts. This paper reviews the available evaluated HALEU fuel benchmarks and some of the potentially relevant benchmarks for fresh highly enriched uranium. It then introduces experimental data for HALEU fuel irradiated at Idaho National Laboratory, from relatively recent irradiation programs at the Advanced Test Reactor. Such data should be evaluated and, if valuable, collected into detailed benchmark specifications to meet the needs of HALEU-based reactor designers.
{"title":"A Review of Candidates for a Validation Data Set for High-Assay Low-Enrichment Uranium Fuels","authors":"M. DeHart, J. Bess, G. Ilas","doi":"10.3390/jne4030038","DOIUrl":"https://doi.org/10.3390/jne4030038","url":null,"abstract":"Many advanced reactor concept designs rely on high-assay low-enriched uranium (HALEU) fuel, enriched up to approximately 19.75% 235U by weight. Efforts are underway by the US government to increase HALEU production in the United States to meet anticipated needs. However, very few data exist for validation of computational models that include HALEU, beyond a few fresh fuel benchmark specifications in the International Reactor Physics Experiment Evaluation Project. Nevertheless, there are other data with potential value available for developing into quality benchmarks for use in data- and software-validation efforts. This paper reviews the available evaluated HALEU fuel benchmarks and some of the potentially relevant benchmarks for fresh highly enriched uranium. It then introduces experimental data for HALEU fuel irradiated at Idaho National Laboratory, from relatively recent irradiation programs at the Advanced Test Reactor. Such data should be evaluated and, if valuable, collected into detailed benchmark specifications to meet the needs of HALEU-based reactor designers.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81215060","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}
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":null,"pages":null},"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":null,"pages":null},"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":null,"pages":null},"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":null,"pages":null},"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}
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