Abstract The ASME Journal of Nuclear Engineering and Radiation Science (NERS) (https://asmedigitalcollection.asme.org/nuclearengineering) celebrates its 10th year from the date of establishing (the Journal was established in 2014) and 9th year from our first issue (published in January of 2015). On this occasion, I would like to congratulate all members of our Journal Board, reviewers (their names are listed at the end of this greeting), authors, and readers with these dates! Also, our Journal, finally, have received a Journal Impact Factor in 2023.
{"title":"Editorial Message to Journal Readers","authors":"Igor Pioro","doi":"10.1115/1.4063758","DOIUrl":"https://doi.org/10.1115/1.4063758","url":null,"abstract":"Abstract The ASME Journal of Nuclear Engineering and Radiation Science (NERS) (https://asmedigitalcollection.asme.org/nuclearengineering) celebrates its 10th year from the date of establishing (the Journal was established in 2014) and 9th year from our first issue (published in January of 2015). On this occasion, I would like to congratulate all members of our Journal Board, reviewers (their names are listed at the end of this greeting), authors, and readers with these dates! Also, our Journal, finally, have received a Journal Impact Factor in 2023.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"39 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135804191","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}
Abstract LOTUS reactor core is a small modular lead-cooled fast reactor with designed power of 200 MWth under development at VNU University of Science, Hanoi for a floating nuclear power plant application. For that purpose, advanced passive safety features and no refuelling requirement are the priorities in the core design process. To endure the continuous operation over a long lifetime, the start-up core exhibits excess reactivity to cover the reactivity loss due to burnup. The reactivity control system includes burnable poison and absorber rods and layers made of B4C which are employed in the reactor to minimize the excess reactivity of the core to about 1 $ to enhance the safety features of the core. The burnable poison is fixed inside the reactor while absorber rods/absorber layers were withdrawn or inserted in sequence to achieve the required excess reactivity of about 700 pcm. The reactivity control was arranged into ten steps to achieve the operating time of 15 effective full-power years without refuelling. Good neutronics behaviour of the core was observed with negative fuel temperature coefficient and coolant void reactivity and maximum radial power peaking factor of 1.32. However, a quite large residual absorption caused by fixed burnable poison inside fuel assemblies was revealed. In further study, to increase the neutron absorption efficiency of burnable poison in the fast spectrum as well as the reactor lifetime, a neutron moderator will be considered to add into the burnable poison rods.
{"title":"Feasibility Study on the Application of Boron Carbide for Long Term Reactivity Control in the LOTUS Small Fast Reactor","authors":"Thanh Mai Vu, Thi Hong Bui, Le Quang Linh Tran","doi":"10.1115/1.4063739","DOIUrl":"https://doi.org/10.1115/1.4063739","url":null,"abstract":"Abstract LOTUS reactor core is a small modular lead-cooled fast reactor with designed power of 200 MWth under development at VNU University of Science, Hanoi for a floating nuclear power plant application. For that purpose, advanced passive safety features and no refuelling requirement are the priorities in the core design process. To endure the continuous operation over a long lifetime, the start-up core exhibits excess reactivity to cover the reactivity loss due to burnup. The reactivity control system includes burnable poison and absorber rods and layers made of B4C which are employed in the reactor to minimize the excess reactivity of the core to about 1 $ to enhance the safety features of the core. The burnable poison is fixed inside the reactor while absorber rods/absorber layers were withdrawn or inserted in sequence to achieve the required excess reactivity of about 700 pcm. The reactivity control was arranged into ten steps to achieve the operating time of 15 effective full-power years without refuelling. Good neutronics behaviour of the core was observed with negative fuel temperature coefficient and coolant void reactivity and maximum radial power peaking factor of 1.32. However, a quite large residual absorption caused by fixed burnable poison inside fuel assemblies was revealed. In further study, to increase the neutron absorption efficiency of burnable poison in the fast spectrum as well as the reactor lifetime, a neutron moderator will be considered to add into the burnable poison rods.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"46 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135918087","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}
Robert David, Blake Mitchell, Justin Spencer, Norm Lair, Sergei Petoukhov
Abstract The properties and behaviour of corium are important factors in the progression and ultimate consequences of a severe nuclear reactor accident. Canadian Nuclear Laboratories (CNL) is developing a cold crucible facility to create and study small quantities of CANDU corium and, potentially, other molten materials. This paper describes the facility and presents preliminary results with non-radioactive charges and with radioactive, near-prototypical charges.
{"title":"Cold Crucible Facility for Severe Accident Research at CNL","authors":"Robert David, Blake Mitchell, Justin Spencer, Norm Lair, Sergei Petoukhov","doi":"10.1115/1.4063673","DOIUrl":"https://doi.org/10.1115/1.4063673","url":null,"abstract":"Abstract The properties and behaviour of corium are important factors in the progression and ultimate consequences of a severe nuclear reactor accident. Canadian Nuclear Laboratories (CNL) is developing a cold crucible facility to create and study small quantities of CANDU corium and, potentially, other molten materials. This paper describes the facility and presents preliminary results with non-radioactive charges and with radioactive, near-prototypical charges.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"89 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135254600","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}
Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model to properly account for the radial leakage effects of the reactor. The leakage cross section and other consistent neutronics parameters are generated with the Monte Carlo code Serpent using high-fidelity three-dimensional (3D) models. The accuracy of the 1D consistent model is verified by the reference solution from the Monte Carlo model on the Molten Salt Reactor Experiment (MSRE) configuration. The 1D consistent model successfully reproduced the integrated flux from the 3D model and the reactor multiplication factor keff with the error in the range of 95 to 397 pcm (per cent mille), depending on discretized energy group structures. The developed model is also extended to estimate the reactivity loss due to fuel circulation in MSRE. The estimate of reactivity loss in dynamics analysis is in great agreement with the experimental data. This model functions as the first step in the development of a 1D fully neutronics/thermal-hydraulics coupled model for short- and long-term MSRE transient analysis.
{"title":"A Consistent One-Dimensional Multigroup Diffusion Model for Molten Salt Reactor Neutronics Calculations","authors":"Mohamed Elhareef, Zeyun Wu, Massimiliano Fratoni","doi":"10.3390/jne4040041","DOIUrl":"https://doi.org/10.3390/jne4040041","url":null,"abstract":"Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model to properly account for the radial leakage effects of the reactor. The leakage cross section and other consistent neutronics parameters are generated with the Monte Carlo code Serpent using high-fidelity three-dimensional (3D) models. The accuracy of the 1D consistent model is verified by the reference solution from the Monte Carlo model on the Molten Salt Reactor Experiment (MSRE) configuration. The 1D consistent model successfully reproduced the integrated flux from the 3D model and the reactor multiplication factor keff with the error in the range of 95 to 397 pcm (per cent mille), depending on discretized energy group structures. The developed model is also extended to estimate the reactivity loss due to fuel circulation in MSRE. The estimate of reactivity loss in dynamics analysis is in great agreement with the experimental data. This model functions as the first step in the development of a 1D fully neutronics/thermal-hydraulics coupled model for short- and long-term MSRE transient analysis.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"5 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135351347","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}
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":"20 1","pages":"0"},"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":"31 1","pages":""},"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":"26 1","pages":""},"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":"14 1","pages":""},"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":"8 1","pages":""},"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":"55 1","pages":""},"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}
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