In discussions on the nuclear restart policy in Japan, trust among stakeholders in nuclear power is often raised as an issue. However, the discussion has progressed without analyzing the trust demanded by the Japanese public. A questionnaire survey was conducted to analyze the level of trust that 4,700 Japanese citizens (100 from each of the 47 prefectures) have in nuclear stakeholders by breaking trust down into three elements: “Perceived Competence,” “Perceived Motivation,” and “Salient Value Similarity” (SVS), which have been reported in the literature. In the survey, nine stakeholders were targeted: the Nuclear Regulation Authority (NRA); the Ministry of Economy, Trade and Industry (METI); the courts; the Diet; host municipalities; nuclear power operators and manufacturers (OMs); experts; research and international organizations (RIOs); and the mass media. The analysis showed that the Japanese public is particularly interested in SVS with “nuclear power operators and manufacturers” among the various nuclear stakeholders. The SVS of “nuclear power operators and manufacturers” was highly correlated with that of organizations promoting nuclear power, suggesting that the public tends to perceive these organizations as sharing similar values. These results offer suggestions for actions that should be taken by each nuclear stakeholder to restore and improve trust.
{"title":"Analysis of Stakeholders' Trust Level in the Nuclear Energy Domain in Japan","authors":"Kyohei Yoshinaga, Masaki Onodera, Kosuke Shirai, Nobuaki Yoshizawa, Harunaga Yanagawa, Takamichi Kito","doi":"10.1115/1.4064778","DOIUrl":"https://doi.org/10.1115/1.4064778","url":null,"abstract":"\u0000 In discussions on the nuclear restart policy in Japan, trust among stakeholders in nuclear power is often raised as an issue. However, the discussion has progressed without analyzing the trust demanded by the Japanese public. A questionnaire survey was conducted to analyze the level of trust that 4,700 Japanese citizens (100 from each of the 47 prefectures) have in nuclear stakeholders by breaking trust down into three elements: “Perceived Competence,” “Perceived Motivation,” and “Salient Value Similarity” (SVS), which have been reported in the literature. In the survey, nine stakeholders were targeted: the Nuclear Regulation Authority (NRA); the Ministry of Economy, Trade and Industry (METI); the courts; the Diet; host municipalities; nuclear power operators and manufacturers (OMs); experts; research and international organizations (RIOs); and the mass media. The analysis showed that the Japanese public is particularly interested in SVS with “nuclear power operators and manufacturers” among the various nuclear stakeholders. The SVS of “nuclear power operators and manufacturers” was highly correlated with that of organizations promoting nuclear power, suggesting that the public tends to perceive these organizations as sharing similar values. These results offer suggestions for actions that should be taken by each nuclear stakeholder to restore and improve trust.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140451283","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}
Thanh Mai Vu, Le Quang Linh Tran, Thi Hong Bui, Nhu Viet Ha Pham
A small modular lead-cooled fast spectrum core concept called ALMANAR designed to produce 45 MWth power for 22 years operating without refuelling was proposed in a previous study. The neutronics investigation showed its excellent inherent safety features. It could be considered as a candidate for future electricity source for the near future. It is noteworthy that the target accuracy for eigenvalue calculation for keff regardless of spectrum is set to 300 pcm. However, findings in this analysis revealed that the keff uncertainty was larger for the recently released nuclear data libraries (about 800 pcm), mostly from 235U capture cross section (624 pcm) in the case of ENDF/B-VIII.0 and 238U inelastic scattering cross section (437 pcm) in the case of JENDL-5. Selected kinetic parameters of the ALMANAR core and their uncertainty were also evaluated and analysed. No major impact on the total ßeff, leffeff and λeff simulation results was found. In order to improve the reliability of criticality calculations of the lead-cooled small fast reactor, the accuracy of capture and fission cross section of 235,238U, the capture cross section of 10B and the elastic scattering cross section of 208Pb at the fast energy range of ENDF/B-VIII.0 should be improved. Furthermore, the inelastic scattering and capture cross section of 238U, fission and capture cross section of 235U and the capture cross section of 10B of JENDL-5 should also be improved.
{"title":"Calculation And Uncertainty Analysis of Core Parameters of ALMANAR Reactor Using New Nuclear Data Libraries","authors":"Thanh Mai Vu, Le Quang Linh Tran, Thi Hong Bui, Nhu Viet Ha Pham","doi":"10.1115/1.4064780","DOIUrl":"https://doi.org/10.1115/1.4064780","url":null,"abstract":"\u0000 A small modular lead-cooled fast spectrum core concept called ALMANAR designed to produce 45 MWth power for 22 years operating without refuelling was proposed in a previous study. The neutronics investigation showed its excellent inherent safety features. It could be considered as a candidate for future electricity source for the near future. It is noteworthy that the target accuracy for eigenvalue calculation for keff regardless of spectrum is set to 300 pcm. However, findings in this analysis revealed that the keff uncertainty was larger for the recently released nuclear data libraries (about 800 pcm), mostly from 235U capture cross section (624 pcm) in the case of ENDF/B-VIII.0 and 238U inelastic scattering cross section (437 pcm) in the case of JENDL-5. Selected kinetic parameters of the ALMANAR core and their uncertainty were also evaluated and analysed. No major impact on the total ßeff, leffeff and λeff simulation results was found. In order to improve the reliability of criticality calculations of the lead-cooled small fast reactor, the accuracy of capture and fission cross section of 235,238U, the capture cross section of 10B and the elastic scattering cross section of 208Pb at the fast energy range of ENDF/B-VIII.0 should be improved. Furthermore, the inelastic scattering and capture cross section of 238U, fission and capture cross section of 235U and the capture cross section of 10B of JENDL-5 should also be improved.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140452180","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}
Jason Schulthess, Katelyn Baird, Philip Petersen, Daniele Salvato, Hakan Ozaltun, William Hanson, Nicholas Ullum, Jeffrey J. Giglio, James I. Cole
A plate-type nuclear fuel consisting of a solid monolithic foil of U-10wt%Mo is under development for use in the United States' high performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate post-irradiation mechanical properties of the commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo samples to generate the fuel material properties, including the modulus of elasticity and the bending strength. Although the material behaves in a brittle manner due to the accumulated porosity, a general trend of strength and modulus reduction were found as fission density increases. The data produced was evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.
{"title":"Mechanical Properties of Irradiated U-10wt%Mo Alloy Degraded by Porosity Development","authors":"Jason Schulthess, Katelyn Baird, Philip Petersen, Daniele Salvato, Hakan Ozaltun, William Hanson, Nicholas Ullum, Jeffrey J. Giglio, James I. Cole","doi":"10.1115/1.4064779","DOIUrl":"https://doi.org/10.1115/1.4064779","url":null,"abstract":"\u0000 A plate-type nuclear fuel consisting of a solid monolithic foil of U-10wt%Mo is under development for use in the United States' high performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate post-irradiation mechanical properties of the commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo samples to generate the fuel material properties, including the modulus of elasticity and the bending strength. Although the material behaves in a brittle manner due to the accumulated porosity, a general trend of strength and modulus reduction were found as fission density increases. The data produced was evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140450514","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}
Canadian Nuclear Laboratories is collaborating in the Joint European Canadian Chinese Development of Small Modular Reactor Technology (ECC-SMART) project to understand the corrosion behavior of the most promising candidate materials for a future supercritical water cooled - small modular reactor (SCW-SMR). To support this aim and the project's requirements, the present study develops a costing method for assessing the impact of corrosion in a power generation cost model. This cost model builds on a methodological study of various corrosion engineering economics topics in nuclear power generation, such as the expected fuel cladding corrosion phenomena in an SCWR concept and estimating the main corrosion costs categories. This understanding is incorporated in a power generation cost model that applies the revenue requirements approach to life cycle costing (LCC). The LCC includes the main corrosion cost categories and a reliability factor used in assessing power generation costs, the costing of chemical species for controlling corrosion, and the present worth of revenue requirements. The method and model, therefore, provide a framework for understanding the kind of information available and needed for taking economical preventative corrosion measures for the current generation of water-cooled reactors and advanced reactors, such as the SCWR.
{"title":"Revenue Requirements Method for Assessing the Cost Impact of Fuel Cladding Corrosion in a Supercritical Water-Cooled Small Modular Reactor: A Methodological Review on Life Cycle Costing Corrosion","authors":"A. D. Mendoza España","doi":"10.1115/1.4064639","DOIUrl":"https://doi.org/10.1115/1.4064639","url":null,"abstract":"\u0000 Canadian Nuclear Laboratories is collaborating in the Joint European Canadian Chinese Development of Small Modular Reactor Technology (ECC-SMART) project to understand the corrosion behavior of the most promising candidate materials for a future supercritical water cooled - small modular reactor (SCW-SMR). To support this aim and the project's requirements, the present study develops a costing method for assessing the impact of corrosion in a power generation cost model. This cost model builds on a methodological study of various corrosion engineering economics topics in nuclear power generation, such as the expected fuel cladding corrosion phenomena in an SCWR concept and estimating the main corrosion costs categories. This understanding is incorporated in a power generation cost model that applies the revenue requirements approach to life cycle costing (LCC). The LCC includes the main corrosion cost categories and a reliability factor used in assessing power generation costs, the costing of chemical species for controlling corrosion, and the present worth of revenue requirements. The method and model, therefore, provide a framework for understanding the kind of information available and needed for taking economical preventative corrosion measures for the current generation of water-cooled reactors and advanced reactors, such as the SCWR.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140477446","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) based on high temperature gas-cooled reactor (HTGR) technology are being developed for providing high-temperature process heat and high-efficiency (>40%) electrical power generation. However, most of the HTGR-SMR concepts require high assay low enriched uranium (HALEU) fuel, with enrichments typically above 10 wt% 235U/U, to get sufficiently high burnup levels and fuel lifetime. The need for HALEU is primarily a consequence of the low volumetric density of fissionable material in tri-structural isotropic (TRISO) fuel particles, and also the use of graphite as a moderator/reflector. A previous study has shown that a modified prismatic HTGR fuel assembly with hydrogen-based moderator (7LiH) and cylindrical fuel elements of 5 wt% 235U/U enriched uranium can greatly reduce fuel consumption of a HTGR. However, such a design concept could lead to positive temperature reactivity coefficients (TRCs), making reactor control more challenging, with reduced passive safety. The purpose of this study is to evaluate variations of the hydrogen-based moderator in this alternative fuel assembly concept to identify configurations that achieve negative TRCs, thus improving passive safety characteristics. Calculation results demonstrate that negative TRCs can be achieved with reduced hydrogen mass such that natural uranium consumption is substantially less than that of the TRISO fuel concept, with comparable or higher core life. This study also explores the option of using 7LiOH and NaOH as hydrogen-based moderators, instead of 7LiH, thus allowing operation at higher temperatures, where hydrogen TRCs are lower.
{"title":"Evaluating Hydrogen-based Moderators in High-Temperature Gas-cooled Reactors with 5 wt% Enriched Uranium Annular Fuel Rods","authors":"D. Wojtaszek, B. Bromley","doi":"10.1115/1.4064581","DOIUrl":"https://doi.org/10.1115/1.4064581","url":null,"abstract":"\u0000 Small Modular Reactors (SMRs) based on high temperature gas-cooled reactor (HTGR) technology are being developed for providing high-temperature process heat and high-efficiency (>40%) electrical power generation. However, most of the HTGR-SMR concepts require high assay low enriched uranium (HALEU) fuel, with enrichments typically above 10 wt% 235U/U, to get sufficiently high burnup levels and fuel lifetime. The need for HALEU is primarily a consequence of the low volumetric density of fissionable material in tri-structural isotropic (TRISO) fuel particles, and also the use of graphite as a moderator/reflector. A previous study has shown that a modified prismatic HTGR fuel assembly with hydrogen-based moderator (7LiH) and cylindrical fuel elements of 5 wt% 235U/U enriched uranium can greatly reduce fuel consumption of a HTGR. However, such a design concept could lead to positive temperature reactivity coefficients (TRCs), making reactor control more challenging, with reduced passive safety. The purpose of this study is to evaluate variations of the hydrogen-based moderator in this alternative fuel assembly concept to identify configurations that achieve negative TRCs, thus improving passive safety characteristics. Calculation results demonstrate that negative TRCs can be achieved with reduced hydrogen mass such that natural uranium consumption is substantially less than that of the TRISO fuel concept, with comparable or higher core life. This study also explores the option of using 7LiOH and NaOH as hydrogen-based moderators, instead of 7LiH, thus allowing operation at higher temperatures, where hydrogen TRCs are lower.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140493022","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}
R. A. P. Dwijayanto, Harun Ardiansyah, A. W. Harto
Thermal molten salt reactors can be designed in many configurations. This paper investigates the optimal geometry of a one fluid molten salt reactor (OFMSR) in a virtual one-and-half fluid configuration with a fixed fuel salt volume. Two primary configurations were studied, axial blanket (three models) and radial blanket (two models). Neutronic calculations were performed using MCNP6.2 and Serpent-2 reactor physics codes with ENDF/B-VII.0 continuous neutron library. The analysis comprises criticality calculation, temperature coefficient of reactivity (TCR), breeding ratio (BR), and kinetic parameters. The results imply a good agreement between MCNP and Serpent calculations. TCR values show a different pattern between axial and radial blanket configuration. Whilst the correlation between TCR and BR is inversely correlated in axial blanket, it is linear in radial blanket configuration. Overall, radial blanket configuration seemed to show better neutronic performance than axial blanket configuration, with comparably strong negative TCR and large BR.
{"title":"Verification and Geometry Optimization of a One Fluid Molten Salt Reactor (OFMSR) with Fixed Volume","authors":"R. A. P. Dwijayanto, Harun Ardiansyah, A. W. Harto","doi":"10.1115/1.4064465","DOIUrl":"https://doi.org/10.1115/1.4064465","url":null,"abstract":"\u0000 Thermal molten salt reactors can be designed in many configurations. This paper investigates the optimal geometry of a one fluid molten salt reactor (OFMSR) in a virtual one-and-half fluid configuration with a fixed fuel salt volume. Two primary configurations were studied, axial blanket (three models) and radial blanket (two models). Neutronic calculations were performed using MCNP6.2 and Serpent-2 reactor physics codes with ENDF/B-VII.0 continuous neutron library. The analysis comprises criticality calculation, temperature coefficient of reactivity (TCR), breeding ratio (BR), and kinetic parameters. The results imply a good agreement between MCNP and Serpent calculations. TCR values show a different pattern between axial and radial blanket configuration. Whilst the correlation between TCR and BR is inversely correlated in axial blanket, it is linear in radial blanket configuration. Overall, radial blanket configuration seemed to show better neutronic performance than axial blanket configuration, with comparably strong negative TCR and large BR.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139440686","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}
V. Richet, Sarah Piochaud, Mouna El Alaoui, R. Plana
Nuclear projects often produce and consume a large amount of knowledge. Capitalization on this knowledge constitutes a significant way to increase efficiency on subsequent projects for any stakeholder. In this study, modelling is used as a main approach to support this capitalization. It constitutes, through graphical layout, a more reliable and robust way to transfer information. Moreover, the use of an interconnected set of models enables organizations to break the silos between the disciplines. The approach proposed is based on the declination of existing "on-the-shelf" elements to benefit from previous implementations. The presented example illustrates how, on a nuclear project, engineering processes have been modelled from knowledge of previous projects. These implementations are all interconnected to constitute a self-supporting set of models as a body of knowledge. This approach has enabled significant time and costs savings during project preparatory and initial phases.
{"title":"Use of Modelling as an Enabler for Cross-Topic Knowledge Management and Ontologies to Support Return of Experience and Replicability of Large Nuclear Projects","authors":"V. Richet, Sarah Piochaud, Mouna El Alaoui, R. Plana","doi":"10.1115/1.4064466","DOIUrl":"https://doi.org/10.1115/1.4064466","url":null,"abstract":"\u0000 Nuclear projects often produce and consume a large amount of knowledge. Capitalization on this knowledge constitutes a significant way to increase efficiency on subsequent projects for any stakeholder. In this study, modelling is used as a main approach to support this capitalization. It constitutes, through graphical layout, a more reliable and robust way to transfer information. Moreover, the use of an interconnected set of models enables organizations to break the silos between the disciplines. The approach proposed is based on the declination of existing \"on-the-shelf\" elements to benefit from previous implementations. The presented example illustrates how, on a nuclear project, engineering processes have been modelled from knowledge of previous projects. These implementations are all interconnected to constitute a self-supporting set of models as a body of knowledge. This approach has enabled significant time and costs savings during project preparatory and initial phases.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140511568","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 refers to a broad class of nuclear reactors that use a molten alkali-halide salt as the primary coolant fluid. This paper pertains to thermal spectrum liquid fuel molten fluoride salt reactors with graphite moderator (MSRs), where the molten salt also dissolves the actinide fuel. Xenon isotope 135, 135Xe, is a fission product that is produced during nuclear energy production and it acts as a neutron poison. Due to the circulating nature of the fuel salt in MSRs, there is a qualitative difference in the behavior of 135Xe in an MSR compared to a solid fueled reactor. Some of the 135Xe produced in fission may end up in the pore space of the graphite moderator used in a MSR. This paper examines the transfer and storage of 135Xe in MSR graphite. Prior publications are reviewed, the porosity of the MSR graphite is examined, governing equations are detailed, film layer production and destruction is discussed, the graphite / salt interface is explored, transport pathways are considered, transfer processes are exposited, the effect of charged species is examined, the solubility of noble gases in molten fluoride salts is examined, the mass diffusion coefficient in molten salts is explored, and the calculation of mass transfer coefficients is described.
{"title":"The Transfer of Xenon-135 to Molten Salt Reactor Graphite","authors":"Terry Price, Ondrej Chvala","doi":"10.1115/1.4064464","DOIUrl":"https://doi.org/10.1115/1.4064464","url":null,"abstract":"\u0000 Molten salt reactors refers to a broad class of nuclear reactors that use a molten alkali-halide salt as the primary coolant fluid. This paper pertains to thermal spectrum liquid fuel molten fluoride salt reactors with graphite moderator (MSRs), where the molten salt also dissolves the actinide fuel. Xenon isotope 135, 135Xe, is a fission product that is produced during nuclear energy production and it acts as a neutron poison. Due to the circulating nature of the fuel salt in MSRs, there is a qualitative difference in the behavior of 135Xe in an MSR compared to a solid fueled reactor. Some of the 135Xe produced in fission may end up in the pore space of the graphite moderator used in a MSR. This paper examines the transfer and storage of 135Xe in MSR graphite. Prior publications are reviewed, the porosity of the MSR graphite is examined, governing equations are detailed, film layer production and destruction is discussed, the graphite / salt interface is explored, transport pathways are considered, transfer processes are exposited, the effect of charged species is examined, the solubility of noble gases in molten fluoride salts is examined, the mass diffusion coefficient in molten salts is explored, and the calculation of mass transfer coefficients is described.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139440891","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}
Challenges with safeguarding molten salt reactor (MSR) designs have prompted the search for enhanced safeguards technologies and revised safeguards materials control & accountancy (MC&A) approaches. A molten salt sampling system is a subsystem being developed to help support facility MC&A in future MSRs by removing salt samples from the primary fuel and/or coolant salt loop of an MSR for chemical and isotopic analysis. To consider the safeguards implications of this molten salt sampling system early in the design process, we employed a safeguards by design approach during the development of a prototype molten salt sampling system. Specifically, we identified and tailored a checklist approach to systematically evaluate the design against recognized safeguards and security attributes. This technical brief describes the molten salt sampling system design and operational concept upon which we applied the safeguards by design methodology, conveys the methods we used to employ the safeguards by design approach on the molten salt sampling system design and discusses the preliminary results and design insights gained from this safeguards by design assessment.
{"title":"Technical Brief: Safeguardability Analysis of a Molten Salt Sampling System Design","authors":"M. Harkema, Steven Krahn, Paul Marotta","doi":"10.1115/1.4064343","DOIUrl":"https://doi.org/10.1115/1.4064343","url":null,"abstract":"Challenges with safeguarding molten salt reactor (MSR) designs have prompted the search for enhanced safeguards technologies and revised safeguards materials control & accountancy (MC&A) approaches. A molten salt sampling system is a subsystem being developed to help support facility MC&A in future MSRs by removing salt samples from the primary fuel and/or coolant salt loop of an MSR for chemical and isotopic analysis. To consider the safeguards implications of this molten salt sampling system early in the design process, we employed a safeguards by design approach during the development of a prototype molten salt sampling system. Specifically, we identified and tailored a checklist approach to systematically evaluate the design against recognized safeguards and security attributes. This technical brief describes the molten salt sampling system design and operational concept upon which we applied the safeguards by design methodology, conveys the methods we used to employ the safeguards by design approach on the molten salt sampling system design and discusses the preliminary results and design insights gained from this safeguards by design assessment.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139170134","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 reliable high-temperature molten salt pump is critical for the development of Fluoride-salt-cooled High-temperature Reactors (FHRs). By supporting the rotating journal, the suitable journal bearing can ensure that the high-temperature molten salt pump runs smoothly and efficiently in the high-temperature fluoride salt over a long period of time. However, many bearing candidates served well for only a short period and experienced several issues. Moreover, the molten salt pump journal misalignment or not is a key factor for the molten salt pump's long-term steady running. In the long-term operation, a misalignment in the journal bearing can result in vibrations and excessive wear on the bearing surface of the molten salt pump. The journal bearing dynamic characteristics is a meaningful sign to accurately assess the journal misalignment. Therefore, it is necessary to investigate the detailed journal bearing dynamic behavior under the high-temperature hydrodynamic fluoride salt lubrication conditions for FHR applications. This study's small amplitude vibration is superimposed on a steady-running journal bearing condition. A FORTRAN 90 program has been written for the journal bearing dynamic behavior analysis. The numerical results are validated with experimental data from the literature. The validated program was employed to predict the dynamic coefficients of high-temperature fluoride salt hydrodynamic lubricated journal bearing various Sommerfeld numbers. This study evaluating the journal bearing dynamic coefficients for molten salt pumps provides guidelines that are helpful for designing molten salt primary pumps.
{"title":"Molten Salt Pump Journal-Bearings Dynamic Characteristics Under Hydrodynamic Lubrication Conditions","authors":"Yuqi Liu, Minghui Chen, S. Che, Adam Burak","doi":"10.1115/1.4064336","DOIUrl":"https://doi.org/10.1115/1.4064336","url":null,"abstract":"A reliable high-temperature molten salt pump is critical for the development of Fluoride-salt-cooled High-temperature Reactors (FHRs). By supporting the rotating journal, the suitable journal bearing can ensure that the high-temperature molten salt pump runs smoothly and efficiently in the high-temperature fluoride salt over a long period of time. However, many bearing candidates served well for only a short period and experienced several issues. Moreover, the molten salt pump journal misalignment or not is a key factor for the molten salt pump's long-term steady running. In the long-term operation, a misalignment in the journal bearing can result in vibrations and excessive wear on the bearing surface of the molten salt pump. The journal bearing dynamic characteristics is a meaningful sign to accurately assess the journal misalignment. Therefore, it is necessary to investigate the detailed journal bearing dynamic behavior under the high-temperature hydrodynamic fluoride salt lubrication conditions for FHR applications. This study's small amplitude vibration is superimposed on a steady-running journal bearing condition. A FORTRAN 90 program has been written for the journal bearing dynamic behavior analysis. The numerical results are validated with experimental data from the literature. The validated program was employed to predict the dynamic coefficients of high-temperature fluoride salt hydrodynamic lubricated journal bearing various Sommerfeld numbers. This study evaluating the journal bearing dynamic coefficients for molten salt pumps provides guidelines that are helpful for designing molten salt primary pumps.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139168605","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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