T. Beuthe, Hooman Jazebizadeh, Aleksandar Vasić, A. Pegarkov, T. Kaya, H. Zahlan
Alkali metal heat pipes can be used for core cooling in advanced micro and small modular reactors. This article provides a detailed review of currently available saturation property correlations for potassium that can be utilized in heat pipe simulation models. Using these properties, numerical models will be developed and employed to simulate and compare the performance of heat pipes models to experimental results. A comprehensive comparative overview of the best available potassium saturation property correlations developed over the past century has been assembled. The results show most of the potassium property correlations needed to model a heat pipe are relatively well defined and recommendations for their use can be provided, but the findings also suggest a significant disparity in some cases.
{"title":"Development of Analysis Tools for Heat Pipes Used in the Core Cooling of Small Modular Reactors: Potassium Property Correlations","authors":"T. Beuthe, Hooman Jazebizadeh, Aleksandar Vasić, A. Pegarkov, T. Kaya, H. Zahlan","doi":"10.1115/1.4062752","DOIUrl":"https://doi.org/10.1115/1.4062752","url":null,"abstract":"\u0000 Alkali metal heat pipes can be used for core cooling in advanced micro and small modular reactors. This article provides a detailed review of currently available saturation property correlations for potassium that can be utilized in heat pipe simulation models. Using these properties, numerical models will be developed and employed to simulate and compare the performance of heat pipes models to experimental results. A comprehensive comparative overview of the best available potassium saturation property correlations developed over the past century has been assembled. The results show most of the potassium property correlations needed to model a heat pipe are relatively well defined and recommendations for their use can be provided, but the findings also suggest a significant disparity in some cases.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90890359","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}
Modeling and simulation lie at the heart of the design process of any nuclear application. An accurate representation of the radiation environment ensures not only the feasibility of new technologies, but it also aids in operation, maintenance, and even decommissioning. With increasingly complex designs, high-fidelity models have become a necessity for design maturity. McCAD has been under development for many years at Karlsruhe Institute of Technology (KIT) to facilitate the process of generating suitable models for nuclear analyses. In this paper, an overview of the major advances in the new version of the code is presented. A novel conversion algorithm has proven to be robust in significantly reducing the processing time to generate radiation transport models, making it easier to iterate on design details. A first-of-a-kind capability to generate hierarchical void cells is also discussed with preliminary analysis showing performance gains for particle tracking.
{"title":"A Novel Algorithm for CAD to CSG Conversion in McCAD","authors":"M. Harb, D. Leichtle, U. Fischer","doi":"10.3390/jne4020031","DOIUrl":"https://doi.org/10.3390/jne4020031","url":null,"abstract":"Modeling and simulation lie at the heart of the design process of any nuclear application. An accurate representation of the radiation environment ensures not only the feasibility of new technologies, but it also aids in operation, maintenance, and even decommissioning. With increasingly complex designs, high-fidelity models have become a necessity for design maturity. McCAD has been under development for many years at Karlsruhe Institute of Technology (KIT) to facilitate the process of generating suitable models for nuclear analyses. In this paper, an overview of the major advances in the new version of the code is presented. A novel conversion algorithm has proven to be robust in significantly reducing the processing time to generate radiation transport models, making it easier to iterate on design details. A first-of-a-kind capability to generate hierarchical void cells is also discussed with preliminary analysis showing performance gains for particle tracking.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-06-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79217950","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}
M. Khalaquzzaman, Seung Jun Lee, Muhammed Mufazzal Hossen
Quality and safety are intensely related and go hand in hand. Quality of the safety-grade equipment is very important for the safety of a nuclear power plant (NPP) and achieving production goals. During manufacturing of plant components or equipment, deviation from the design might occur at different stages of manufacturing for various reasons, such as a lack of skilled manpower, deviation of materials, human errors, malfunction of equipment, violation of manufacturing procedure, etc. These deviations can be assessed cautiously and taken into consideration in the final safety analysis report (FSAR) before issuing an operating license. In this paper, we propose a Bayesian belief network for quality assessment of safety class equipment of NPPs with a few examples. The proposed procedure is a holistic approach for estimation of equipment failure probability considering manufacturing deviations and errors. Case studies for safety-class dry transformers and reactor pressurizers employing the proposed method are also presented in this article. This study provides insights for probabilistic safety assessment engineers and nuclear plant regulators for improved assessment of NPP safety.
{"title":"Reliability Assessment of NPP Safety Class Equipment Considering the Manufacturing Quality Assurance Process","authors":"M. Khalaquzzaman, Seung Jun Lee, Muhammed Mufazzal Hossen","doi":"10.3390/jne4020030","DOIUrl":"https://doi.org/10.3390/jne4020030","url":null,"abstract":"Quality and safety are intensely related and go hand in hand. Quality of the safety-grade equipment is very important for the safety of a nuclear power plant (NPP) and achieving production goals. During manufacturing of plant components or equipment, deviation from the design might occur at different stages of manufacturing for various reasons, such as a lack of skilled manpower, deviation of materials, human errors, malfunction of equipment, violation of manufacturing procedure, etc. These deviations can be assessed cautiously and taken into consideration in the final safety analysis report (FSAR) before issuing an operating license. In this paper, we propose a Bayesian belief network for quality assessment of safety class equipment of NPPs with a few examples. The proposed procedure is a holistic approach for estimation of equipment failure probability considering manufacturing deviations and errors. Case studies for safety-class dry transformers and reactor pressurizers employing the proposed method are also presented in this article. This study provides insights for probabilistic safety assessment engineers and nuclear plant regulators for improved assessment of NPP safety.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-06-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89622535","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Shmelev, N. Geraskin, V. Apse, V. Glebov, E. Kulikov, A. Krasnoborodko
This paper presents the results obtained from numerical evaluations for the possibility of large-scale 238Pu production in the light-water VVER-1000 reactor and the reactivity effect caused by the loss-of-coolant accident in the central fuel assembly of the reactor core. This fuel assembly containing the Np–Am-component of minor actinides was placed in the center of the reactor core and intended for intense production of 238Pu. Optimal conditions were found for large-scale production of plutonium with an isotope composition suitable for application in radioisotope thermoelectric generators. The reactivity effect from the loss-of-coolant accident in the central Np–Am fuel assembly was evaluated, and the perturbation theory was used to determine the contributions of some neutron processes (leakage, absorption, and moderation) to the total variation of the effective neutron multiplication factor.
{"title":"Application of Np–Am Mixture in Production of 238Pu in a VVER-1000 Reactor and the Reactivity Effect Caused by Loss-of-Coolant Accident in the Central Np–Am Fuel Assembly","authors":"A. Shmelev, N. Geraskin, V. Apse, V. Glebov, E. Kulikov, A. Krasnoborodko","doi":"10.3390/jne4020029","DOIUrl":"https://doi.org/10.3390/jne4020029","url":null,"abstract":"This paper presents the results obtained from numerical evaluations for the possibility of large-scale 238Pu production in the light-water VVER-1000 reactor and the reactivity effect caused by the loss-of-coolant accident in the central fuel assembly of the reactor core. This fuel assembly containing the Np–Am-component of minor actinides was placed in the center of the reactor core and intended for intense production of 238Pu. Optimal conditions were found for large-scale production of plutonium with an isotope composition suitable for application in radioisotope thermoelectric generators. The reactivity effect from the loss-of-coolant accident in the central Np–Am fuel assembly was evaluated, and the perturbation theory was used to determine the contributions of some neutron processes (leakage, absorption, and moderation) to the total variation of the effective neutron multiplication factor.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77731252","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}
Xingyu Zhao, Minyun Liu, Rongyi Cui, Shanfang Huang, Kan Wang, Chuan Lu
Abstract This study analyzed an yttrium hydride (YH2) moderated supercritical carbon dioxide cooled reactor loaded with pin-type, beryllium oxide diluted oxide fuel elements to reduce the critical enrichment. The impact of the YH2 on the coolant void reactivity was studied along with a moderator zoning scheme to flatten the radial power distribution. The YH2 was added as hexagonal moderating rods at the center of the fuel assemblies. The core was modeled using the continuous-energy Reactor Monte Carlo code (RMC) with the on-the-fly cross sections treatment. The results showed that the YH2 moderator increased the thermal fission and reduced the critical enrichment of the core with the same diluent volume fraction by more than 30%. The YH2 moderator significantly softened the neutron energy spectrum and reduced the neutron leakage upon core voiding, resulting in both a weaker positive spectral reactivity feedback and a weaker negative leakage reactivity feedback during core depressurization. For an UO2-loaded core, the YH2 gave a lower negative coolant void reactivity, while for a mixed oxide fuel (MOX)-loaded core with diluent volume fractions smaller than 35%, the spectral feedback was more important and the YH2 strongly reduced the positive coolant void reactivity to less than $1. Arranging the YH2 in the peripheral assemblies reduced the radial power peaking factor to 1.319. The study shows that the YH2 moderator can reduce the critical enrichment, make the core less sensitive to voiding, and can flatten the radial power distribution of a single-enrichment core through moderator zoning.
{"title":"Optimized Moderator Design and Analysis of a Pin-Type Supercritical Carbon Dioxide Reactor Based on Reactor Monte Carlo Code","authors":"Xingyu Zhao, Minyun Liu, Rongyi Cui, Shanfang Huang, Kan Wang, Chuan Lu","doi":"10.1115/1.4056772","DOIUrl":"https://doi.org/10.1115/1.4056772","url":null,"abstract":"Abstract This study analyzed an yttrium hydride (YH2) moderated supercritical carbon dioxide cooled reactor loaded with pin-type, beryllium oxide diluted oxide fuel elements to reduce the critical enrichment. The impact of the YH2 on the coolant void reactivity was studied along with a moderator zoning scheme to flatten the radial power distribution. The YH2 was added as hexagonal moderating rods at the center of the fuel assemblies. The core was modeled using the continuous-energy Reactor Monte Carlo code (RMC) with the on-the-fly cross sections treatment. The results showed that the YH2 moderator increased the thermal fission and reduced the critical enrichment of the core with the same diluent volume fraction by more than 30%. The YH2 moderator significantly softened the neutron energy spectrum and reduced the neutron leakage upon core voiding, resulting in both a weaker positive spectral reactivity feedback and a weaker negative leakage reactivity feedback during core depressurization. For an UO2-loaded core, the YH2 gave a lower negative coolant void reactivity, while for a mixed oxide fuel (MOX)-loaded core with diluent volume fractions smaller than 35%, the spectral feedback was more important and the YH2 strongly reduced the positive coolant void reactivity to less than $1. Arranging the YH2 in the peripheral assemblies reduced the radial power peaking factor to 1.319. The study shows that the YH2 moderator can reduce the critical enrichment, make the core less sensitive to voiding, and can flatten the radial power distribution of a single-enrichment core through moderator zoning.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-05-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135195823","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Saraswat, R. Bhattacharyay, P. Chaudhuri, S. Gedupudi
Liquid metals (LMs) exhibit several key characteristics justifying their utilization as coolants/breeders for nuclear fusion reactors and advanced fission reactors. In a fusion reactor, LMs confront a large flow retarding MHD force, imposing significant demands on pumping power and designs of ancillary systems. Corrosion of structural materials and coolant chemistry control are vital issues common to both fusion and fission reactors employing liquid lead (Pb) and its alloys. To address these concerns, technological solutions such as Flow Channel Inserts and corrosion resistant coatings are being investigated to provide a chemical/electrical isolation between LM and structural material. In this study, three prototype geometries (circular, square and 90 bend) of steel-insulator-steel FCIs are fabricated and an electrical insulation characterization is performed over a temperature range of 100C - 600C. Welding trials and pressure tests are performed to validate the electrical and mechanical integrity over typical fusion reactor operational regime. This paper presents detailed fabrication aspects along with quantitative estimations of insulation filling density, electrical insulation performance and, for the first time, a systematic study of insulation degradation owing to combined effects of TIG welding, pressure and machining operations. Critical details derived from metallurgical examinations and destructive tests are also presented. From implementation perspective towards LFRs, a feasibility assessment of a-Al2O3/AlPO4 thin film coating deposition on planar and non-planar substrates is performed followed by its mechanical characterizations. Detailed metallurgical analyses are presented to assess Pb ingress after 700 hour exposure to molten Pb alloy.
{"title":"Fabrication Aspects and Performance Characterization of α-Al2O3/ALPO4 Based Sandwich Configuration Flow Channel Inserts and Coatings for High Temperature Liquid Metal Applications","authors":"A. Saraswat, R. Bhattacharyay, P. Chaudhuri, S. Gedupudi","doi":"10.1115/1.4062646","DOIUrl":"https://doi.org/10.1115/1.4062646","url":null,"abstract":"\u0000 Liquid metals (LMs) exhibit several key characteristics justifying their utilization as coolants/breeders for nuclear fusion reactors and advanced fission reactors. In a fusion reactor, LMs confront a large flow retarding MHD force, imposing significant demands on pumping power and designs of ancillary systems. Corrosion of structural materials and coolant chemistry control are vital issues common to both fusion and fission reactors employing liquid lead (Pb) and its alloys. To address these concerns, technological solutions such as Flow Channel Inserts and corrosion resistant coatings are being investigated to provide a chemical/electrical isolation between LM and structural material. In this study, three prototype geometries (circular, square and 90 bend) of steel-insulator-steel FCIs are fabricated and an electrical insulation characterization is performed over a temperature range of 100C - 600C. Welding trials and pressure tests are performed to validate the electrical and mechanical integrity over typical fusion reactor operational regime. This paper presents detailed fabrication aspects along with quantitative estimations of insulation filling density, electrical insulation performance and, for the first time, a systematic study of insulation degradation owing to combined effects of TIG welding, pressure and machining operations. Critical details derived from metallurgical examinations and destructive tests are also presented. From implementation perspective towards LFRs, a feasibility assessment of a-Al2O3/AlPO4 thin film coating deposition on planar and non-planar substrates is performed followed by its mechanical characterizations. Detailed metallurgical analyses are presented to assess Pb ingress after 700 hour exposure to molten Pb alloy.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77085142","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}
Tu Nguyen, E. Patterson, Richard J. K. Taylor, Y. Tseng, C. Waldon
The differences between nuclear fission and fusion have been discussed widely in the literature. However, little has been done to investigate the key differences in safety designs and regulatory requirements between the nuclear reactor types. In this study, an innovative methodology was successfully developed to map nuclear safety features to the fundamental safety principles set out by the nuclear regulators. Three safety cases were assessed in the mapping study, a research fusion reactor (Joint European Torus), a research fission reactor (Tsing Hua Open-pool Reactor) and a commercial fission reactor (Hinkley Point C). The graphical representation allowed a comparative analysis of the safety features and fundamental principles which revealed differences between the hazard profiles of fission and fusion reactors and provided important insights for the creation of a similar map for a future commercial fusion device.
{"title":"Comparative Maps of Safety Features for Fission and Fusion Reactors","authors":"Tu Nguyen, E. Patterson, Richard J. K. Taylor, Y. Tseng, C. Waldon","doi":"10.1115/1.4062643","DOIUrl":"https://doi.org/10.1115/1.4062643","url":null,"abstract":"\u0000 The differences between nuclear fission and fusion have been discussed widely in the literature. However, little has been done to investigate the key differences in safety designs and regulatory requirements between the nuclear reactor types. In this study, an innovative methodology was successfully developed to map nuclear safety features to the fundamental safety principles set out by the nuclear regulators. Three safety cases were assessed in the mapping study, a research fusion reactor (Joint European Torus), a research fission reactor (Tsing Hua Open-pool Reactor) and a commercial fission reactor (Hinkley Point C). The graphical representation allowed a comparative analysis of the safety features and fundamental principles which revealed differences between the hazard profiles of fission and fusion reactors and provided important insights for the creation of a similar map for a future commercial fusion device.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72454669","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}
Small Modular Reactors (SMRs) have gained international attention due to their modular design, small footprint, and lower capital costs for research, development, and construction compared to conventional reactors. Many types of SMRs are being developed in different countries, and regulatory agencies are working on a robust and “harmonized” SMR regulatory framework to ensure safety and protect the environment. However, there are still many details that need to be understood, such as nuclear fuel behavior at high pressures and temperatures (1000 °C), types and levels of radiation exposure during normal operations and accidents, types and volume of nuclear waste, and their proper storage and disposal. Moreover, SMRs' modular design and small size make them suitable for remote locations, such as the Canadian Arctic. However, before introducing this technology, a detailed study of the arctic soil (permafrost) is needed in the context of changing climate. Probabilistic risk assessment (PRA) is a crucial methodology for assessing the safety and reliability of nuclear power plants. Due to multi-module nature of SMRs, cross-unit interactions ( multi-module effects) need to be evaluated as part of the total plant safety assessment. Additionally, given the nature of fuels (low-enriched uranium) and the possible remote location with minimal technical staff, nuclear materials may have a higher probability of diversion. Therefore, nuclear safeguards and material accountancy are essential to prevent nuclear proliferation. This paper discusses the benefits and challenges of deploying different SMR technologies for electricity generation and other applications.
{"title":"Small Modular Reactors: Opportunities and Challenges as Emerging Nuclear Technologies for Power Production","authors":"L. Ghimire, E. Waller","doi":"10.1115/1.4062644","DOIUrl":"https://doi.org/10.1115/1.4062644","url":null,"abstract":"\u0000 Small Modular Reactors (SMRs) have gained international attention due to their modular design, small footprint, and lower capital costs for research, development, and construction compared to conventional reactors. Many types of SMRs are being developed in different countries, and regulatory agencies are working on a robust and “harmonized” SMR regulatory framework to ensure safety and protect the environment. However, there are still many details that need to be understood, such as nuclear fuel behavior at high pressures and temperatures (1000 °C), types and levels of radiation exposure during normal operations and accidents, types and volume of nuclear waste, and their proper storage and disposal. Moreover, SMRs' modular design and small size make them suitable for remote locations, such as the Canadian Arctic. However, before introducing this technology, a detailed study of the arctic soil (permafrost) is needed in the context of changing climate. Probabilistic risk assessment (PRA) is a crucial methodology for assessing the safety and reliability of nuclear power plants. Due to multi-module nature of SMRs, cross-unit interactions ( multi-module effects) need to be evaluated as part of the total plant safety assessment. Additionally, given the nature of fuels (low-enriched uranium) and the possible remote location with minimal technical staff, nuclear materials may have a higher probability of diversion. Therefore, nuclear safeguards and material accountancy are essential to prevent nuclear proliferation. This paper discusses the benefits and challenges of deploying different SMR technologies for electricity generation and other applications.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87101671","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}
Pierrick Agullo, A. Gossard, G. Ranchoux, E. Porcheron, Fabrice Petitot
Several nuclear facilities are currently dismantling in France, namely on CEA's and EDF's sites. The whole decommissioning and dismantling, cutting operations and nuclear waste management can significantly affect worker risks. In this context, this work aims to better assess the internal exposure risks for operations where removal factor and airborne release factor play an essential part in the uncertainties. For that purpose, developing a new method for assessing the risk of internal exposure is necessary to optimize the choice of Personal Protective Equipment during nuclear dismantling. Based on a bibliographic study and feedback, we identified most of the forces and parameters that influence labile contamination removal and particles resuspension. Their high number forced us to highlight the most relevant ones regarding data from the bibliography and worksite reality. Thus, the role of the drying temperature, relative humidity, roughness of the contaminated surface and the swiping pressure were particularly studied on the assessment of the labile contamination and the removal factor. For that, we performed successive wiping operations with cotton wipes on surface on which a simulated labile contamination was deposited to estimate the influence of different parameters on the total labile contamination and the removal factor. Thus, we have highlighted the high influence of roughness on these latter compare to the other identified parameters.
{"title":"Main Parameters Influencing The Level Of Labile Contamination And The Removal Factor","authors":"Pierrick Agullo, A. Gossard, G. Ranchoux, E. Porcheron, Fabrice Petitot","doi":"10.1115/1.4062550","DOIUrl":"https://doi.org/10.1115/1.4062550","url":null,"abstract":"\u0000 Several nuclear facilities are currently dismantling in France, namely on CEA's and EDF's sites. The whole decommissioning and dismantling, cutting operations and nuclear waste management can significantly affect worker risks. In this context, this work aims to better assess the internal exposure risks for operations where removal factor and airborne release factor play an essential part in the uncertainties. For that purpose, developing a new method for assessing the risk of internal exposure is necessary to optimize the choice of Personal Protective Equipment during nuclear dismantling.\u0000 Based on a bibliographic study and feedback, we identified most of the forces and parameters that influence labile contamination removal and particles resuspension. Their high number forced us to highlight the most relevant ones regarding data from the bibliography and worksite reality. Thus, the role of the drying temperature, relative humidity, roughness of the contaminated surface and the swiping pressure were particularly studied on the assessment of the labile contamination and the removal factor. For that, we performed successive wiping operations with cotton wipes on surface on which a simulated labile contamination was deposited to estimate the influence of different parameters on the total labile contamination and the removal factor. Thus, we have highlighted the high influence of roughness on these latter compare to the other identified parameters.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-05-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76852060","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}
Nicholas Dunkle, Alex Wheeler, Jarod Richardson, Sandra Bogetic, Ondrej Chvala, Steven E. Skutnik
Fluid-fueled molten-salt reactors (MSRs) are actively being developed by several companies, with plans to deploy them internationally. The current IAEA inspection tools are largely incompatible with the unique design features of liquid fuel MSRs (e.g., the complex fuel chemistry, circulating fuel inventory, bulk accountancy, and high radiation environment). For these reasons, safeguards for MSRs are seen as challenging and require the development of new techniques. This paper proposes one such technique through the observation of the reactor’s off-gas. Any reactor design using low-enriched uranium will build up plutonium as the fuel undergoes burnup. Plutonium has different fission product yields than uranium. Therefore, a shift in fission product production is expected with fuel evolution. The passive removal of certain gaseous fission products to the off-gas tank of an MSR provides a valuable opportunity for analysis without significant modifications to the design of the system. Uniquely, due to the gaseous nature of the isotopes, beta particle emissions are available for observation. The ratios of these fission product isotopes can, thus, be traced back to the relative amount and types of fissile isotopes in the core. This proposed technique represents an effective safeguards tool for bulk accountancy which, while avoiding being onerous, could be used in concert with other techniques to meet the IAEA’s timeliness goals for the detection of a diversion.
{"title":"Plutonium Signatures in Molten-Salt Reactor Off-Gas Tank and Safeguards Considerations","authors":"Nicholas Dunkle, Alex Wheeler, Jarod Richardson, Sandra Bogetic, Ondrej Chvala, Steven E. Skutnik","doi":"10.3390/jne4020028","DOIUrl":"https://doi.org/10.3390/jne4020028","url":null,"abstract":"Fluid-fueled molten-salt reactors (MSRs) are actively being developed by several companies, with plans to deploy them internationally. The current IAEA inspection tools are largely incompatible with the unique design features of liquid fuel MSRs (e.g., the complex fuel chemistry, circulating fuel inventory, bulk accountancy, and high radiation environment). For these reasons, safeguards for MSRs are seen as challenging and require the development of new techniques. This paper proposes one such technique through the observation of the reactor’s off-gas. Any reactor design using low-enriched uranium will build up plutonium as the fuel undergoes burnup. Plutonium has different fission product yields than uranium. Therefore, a shift in fission product production is expected with fuel evolution. The passive removal of certain gaseous fission products to the off-gas tank of an MSR provides a valuable opportunity for analysis without significant modifications to the design of the system. Uniquely, due to the gaseous nature of the isotopes, beta particle emissions are available for observation. The ratios of these fission product isotopes can, thus, be traced back to the relative amount and types of fissile isotopes in the core. This proposed technique represents an effective safeguards tool for bulk accountancy which, while avoiding being onerous, could be used in concert with other techniques to meet the IAEA’s timeliness goals for the detection of a diversion.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-05-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135718117","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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