This paper discusses permanent deformation of the fuel assembly of pressurized water reactors that can satisfy a drop time requirement of the control rod assembly. During a seismic event, the fuel assemblies and the reactor core exhibit transient vibration. When the fuel assemblies collide, they develop impact forces in the spacer grids along the fuel assemblies. If the impact forces on some spacer grids of the fuel assembly exceed an allowable strength, the buckling occurs in the corresponding spacer grids, followed by permanent deformation. The geometry change of the buckled spacer grids leads to the movement of the guide tube positions in the fuel assembly. Therefore the geometry change is associated with the distortion of the control rod drop path. In this study, the permanent deformation of the spacer grids was measured during the impact tests, and the maximum deviation of the guide tube from the original position was reviewed. Control rod drop tests were also performed to measure the change in drop time according to the deformation of the fuel assembly. Using actuators in the loop test facility, the authors realized the first and second bending mode shapes of the fuel assembly under the flowing water condition. Based on a series of drop tests, the critical fuel assembly deformation was identified that could lead to a violation of an allowable drop time of the control rod.
{"title":"Experimental Investigation of Control Rod Drops into Perturbed Fuel Assemblies","authors":"N. Park, O. Kwon, Y. Yoo, Tae-Suk Kim","doi":"10.1115/1.4062275","DOIUrl":"https://doi.org/10.1115/1.4062275","url":null,"abstract":"\u0000 This paper discusses permanent deformation of the fuel assembly of pressurized water reactors that can satisfy a drop time requirement of the control rod assembly. During a seismic event, the fuel assemblies and the reactor core exhibit transient vibration. When the fuel assemblies collide, they develop impact forces in the spacer grids along the fuel assemblies. If the impact forces on some spacer grids of the fuel assembly exceed an allowable strength, the buckling occurs in the corresponding spacer grids, followed by permanent deformation. The geometry change of the buckled spacer grids leads to the movement of the guide tube positions in the fuel assembly. Therefore the geometry change is associated with the distortion of the control rod drop path. In this study, the permanent deformation of the spacer grids was measured during the impact tests, and the maximum deviation of the guide tube from the original position was reviewed. Control rod drop tests were also performed to measure the change in drop time according to the deformation of the fuel assembly. Using actuators in the loop test facility, the authors realized the first and second bending mode shapes of the fuel assembly under the flowing water condition. Based on a series of drop tests, the critical fuel assembly deformation was identified that could lead to a violation of an allowable drop time of the control rod.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"75 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-04-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83022679","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}
Neutron thermalization leads to the complexity of the scattering cross-section calculation, which influences the accuracy of the neutron transport calculation in the thermal energy range. The higher precision of thermal scattering data is demanded in the small modular reactors (SMRs) design, especially for small-sized PWRs and SCWRs. Additionally, the thermal neutron scattering problems in supercritical water have not yet been solved. In this study, the thermal neutron scattering problems in subcritical water are tested. Based on thermal neutron scattering theory, the GA model and IKE model were analyzed. This work selected the corresponding input parameters, such as the frequency spectrum, the discrete oscillator energy, weight parameters and so on, as well as preliminary studies on how to calculate the thermal scattering data for HinH2O to accomplish the calculation at various temperatures by developing LIPER code. The deviation between the calculated and reference results, which were both obtained by the Monte Carlo code, COSRMC, was below 0.2 pcm. The deviation of the scattering cross-section between the calculation results and reference was below 0.1%, indicating the reasonability of this study’s thermal scattering data calculation.
{"title":"Preliminary Study on the Thermal Neutron Scattering Cross-Section for HinH2O in Small Modular Reactors","authors":"Jun Wu, Yixue Chen","doi":"10.3390/jne4020023","DOIUrl":"https://doi.org/10.3390/jne4020023","url":null,"abstract":"Neutron thermalization leads to the complexity of the scattering cross-section calculation, which influences the accuracy of the neutron transport calculation in the thermal energy range. The higher precision of thermal scattering data is demanded in the small modular reactors (SMRs) design, especially for small-sized PWRs and SCWRs. Additionally, the thermal neutron scattering problems in supercritical water have not yet been solved. In this study, the thermal neutron scattering problems in subcritical water are tested. Based on thermal neutron scattering theory, the GA model and IKE model were analyzed. This work selected the corresponding input parameters, such as the frequency spectrum, the discrete oscillator energy, weight parameters and so on, as well as preliminary studies on how to calculate the thermal scattering data for HinH2O to accomplish the calculation at various temperatures by developing LIPER code. The deviation between the calculated and reference results, which were both obtained by the Monte Carlo code, COSRMC, was below 0.2 pcm. The deviation of the scattering cross-section between the calculation results and reference was below 0.1%, indicating the reasonability of this study’s thermal scattering data calculation.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"108 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74838789","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}
We present measurement results for the flow rate of liquid sodium at temperatures up to 700°C that were obtained with a high temperature prototype of an immersed Eddy Current Flow Meter. The experimental campaign was conducted at the SOLTEC-2 sodium loop at KIT. The main objective of the experiments is the high temperature qualification of the Eddy Current Flow Meter as part of the safety instrumentation of generation IV liquid metal cooled fast reactors. There it is intended to be used for monitoring the flow rate of the coolant and to detect possible blockages of sub assemblies. Due to the large liquid metal volume, the sensor has to be located close to the sub assemblies, therefore measurements from outside of the vessel are not possible and an immersed sensor is required. We demonstrate the successful application of the immersed Eddy Current Flow Meter at such high temperatures and identify the relevant effects with impact on the sensor performance.
{"title":"Eddy Current Flow Meter Measurements in Liquid Sodium at High Temperatures","authors":"N. Krauter, A. Onea, G. Gerbeth, S. Eckert","doi":"10.1115/1.4062239","DOIUrl":"https://doi.org/10.1115/1.4062239","url":null,"abstract":"\u0000 We present measurement results for the flow rate of liquid sodium at temperatures up to 700°C that were obtained with a high temperature prototype of an immersed Eddy Current Flow Meter. The experimental campaign was conducted at the SOLTEC-2 sodium loop at KIT. The main objective of the experiments is the high temperature qualification of the Eddy Current Flow Meter as part of the safety instrumentation of generation IV liquid metal cooled fast reactors. There it is intended to be used for monitoring the flow rate of the coolant and to detect possible blockages of sub assemblies. Due to the large liquid metal volume, the sensor has to be located close to the sub assemblies, therefore measurements from outside of the vessel are not possible and an immersed sensor is required. We demonstrate the successful application of the immersed Eddy Current Flow Meter at such high temperatures and identify the relevant effects with impact on the sensor performance.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"5 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89229692","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}
J. Friconneau, T. Batal, O. David, Chiara Di Paolo, F. Ferlay, S. Gazzotti, L. Giancarli, Christophe Lacroix, J. Martins, Benjamin Michel, J. Travere
The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each PP includes two TBMs. After each 18-month-long plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs with new ones during the ITER long-term shutdown, called long-term maintenance (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace (PI), called pipe forest (PF). TBSs shall be designed so that occupational radiation exposure (ORE) can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER policy. To implement ALARA process requirements, design activities shall consider careful integration investigations starting from the early phase to address all engineering aspects of the replacement sequence. The case study focuses on the PF replacement, in particular the port cell operations. This paper describes the investigations and findings of the ALARA optimisation process implementation in the early engineering phase of the PF.
{"title":"ITER Test Blanket Module—ALARA Investigations for Port Cell Pipe Forest Replacement","authors":"J. Friconneau, T. Batal, O. David, Chiara Di Paolo, F. Ferlay, S. Gazzotti, L. Giancarli, Christophe Lacroix, J. Martins, Benjamin Michel, J. Travere","doi":"10.3390/jne4010022","DOIUrl":"https://doi.org/10.3390/jne4010022","url":null,"abstract":"The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each PP includes two TBMs. After each 18-month-long plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs with new ones during the ITER long-term shutdown, called long-term maintenance (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace (PI), called pipe forest (PF). TBSs shall be designed so that occupational radiation exposure (ORE) can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER policy. To implement ALARA process requirements, design activities shall consider careful integration investigations starting from the early phase to address all engineering aspects of the replacement sequence. The case study focuses on the PF replacement, in particular the port cell operations. This paper describes the investigations and findings of the ALARA optimisation process implementation in the early engineering phase of the PF.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"67 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75411960","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 A new subassembly type passive reactor shutdown device is proposed to expand the diversity and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). The device contains pins with a fuel material that is in a solid state during normal operation but melts and fluidizes during an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident. By rapidly transferring the liquefied device fuel into the lower plenum region of the pins via gravitation alone, the device passively provides high negative reactivity to the core. This study evaluated the nuclear and thermal properties of the device subassembly with metallic fuel to determine the device specifications for proper device operation during ULOF and UTOP accidents. The results of the transient analysis of the ULOF initiating phase in a 750-MWel-class mixed-oxide-fueled SFR core confirmed that a conventional homogeneous core maintains stable cooling of the core before coolant boiling in the driver fuel subassemblies. In contrast, the negative reactivity required to terminate the event by device operation was slightly higher in the low sodium void reactivity core than in the conventional homogeneous core.
{"title":"Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Study on Device Specifications","authors":"Koji Morita, Wei Liu, Tatsumi Arima, Yuji Arita, Isamu Sato, Haruaki Matsuura, Yoshihiro Sekio, Hiroshi Sagara, Masatoshi Kawashima","doi":"10.1115/1.4056854","DOIUrl":"https://doi.org/10.1115/1.4056854","url":null,"abstract":"Abstract A new subassembly type passive reactor shutdown device is proposed to expand the diversity and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). The device contains pins with a fuel material that is in a solid state during normal operation but melts and fluidizes during an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident. By rapidly transferring the liquefied device fuel into the lower plenum region of the pins via gravitation alone, the device passively provides high negative reactivity to the core. This study evaluated the nuclear and thermal properties of the device subassembly with metallic fuel to determine the device specifications for proper device operation during ULOF and UTOP accidents. The results of the transient analysis of the ULOF initiating phase in a 750-MWel-class mixed-oxide-fueled SFR core confirmed that a conventional homogeneous core maintains stable cooling of the core before coolant boiling in the driver fuel subassemblies. In contrast, the negative reactivity required to terminate the event by device operation was slightly higher in the low sodium void reactivity core than in the conventional homogeneous core.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"137 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135747976","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 Betavoltaics are direct conversion energy devices that are ideal for low power and long-lasting, uninterruptable applications. Betavoltaics operate similarly to photovoltaics where a radio isotope irradiates beta particles into a semiconductor p-n junction that converts the kinetic energy into electrical energy. Betavoltaics are limited by their power output from the radio isotope. However, the source power density can be increased by the selection of solid-state substrates. While tritium absorbing substrates can be simulated to estimate tritium absorption levels and surface emission energies, their viability has to be physically evaluated. A state-of-the-art hydrogen loading system developed by our research group was used to evaluate different film types to understand how they perform during the hydrogen/tritium loading process. The hydrogen loading system utilizes the Sievert technique, where the temperature and volume is constant and pressure drop of the system is used to determine hydrogen uptake of a film substrate. The hydrogen loading system procedure was verified using 250 nm thick palladium films at three loading temperatures. Results clearly show uptake of hydrogen by the thin palladium films accurate to the ideal stoichiometric ratio of one hydrogen atom to host palladium atom.
{"title":"Hydrogen Loading System for Thin Films for Betavoltaics","authors":"Darrell Cheu, Thomas Adams, Shripad T. Revankar","doi":"10.1115/1.4056974","DOIUrl":"https://doi.org/10.1115/1.4056974","url":null,"abstract":"Abstract Betavoltaics are direct conversion energy devices that are ideal for low power and long-lasting, uninterruptable applications. Betavoltaics operate similarly to photovoltaics where a radio isotope irradiates beta particles into a semiconductor p-n junction that converts the kinetic energy into electrical energy. Betavoltaics are limited by their power output from the radio isotope. However, the source power density can be increased by the selection of solid-state substrates. While tritium absorbing substrates can be simulated to estimate tritium absorption levels and surface emission energies, their viability has to be physically evaluated. A state-of-the-art hydrogen loading system developed by our research group was used to evaluate different film types to understand how they perform during the hydrogen/tritium loading process. The hydrogen loading system utilizes the Sievert technique, where the temperature and volume is constant and pressure drop of the system is used to determine hydrogen uptake of a film substrate. The hydrogen loading system procedure was verified using 250 nm thick palladium films at three loading temperatures. Results clearly show uptake of hydrogen by the thin palladium films accurate to the ideal stoichiometric ratio of one hydrogen atom to host palladium atom.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"153 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135905066","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}
Additive manufacturing (AM) is a transformational digital manufacturing technology featured with rapidity, customizability, precision, and economy, which is fundamentally altering the way components are designed and manufactured. AM enables the freedom of design, and makes full use of complexity of geometry which "comes for free". Applying AM technology to nuclear industry can yield advanced reactor designs with function and structure matched for the best thermal, fluidic and mechanical performance. In this work, an AM-informed reactor core design with SiC matrix and TRISO particle fuel is proposed and analyzed. The core is an integrated 3D-printed SiC bulk with helical cruciform coolant channels, and the UO2-TRISO fuel particles are dispersed in the bulk. A multi-physics analysis framework for irregular geometry is developed to analyze and further optimize the reactor design. The TRISO particle positions are generated with discrete element method. The Reactor Monte Carlo code (RMC) and the CFD software STAR-CCM+ are used for the neutronic and thermal-fluidic analyses, respectively. RMC simulates the neutron transport to predict the effective multiplication factor and power distribution. STAR-CCM+ calculates the flow and heat transfer in coolant channels and heat conduction in solid matrix with the power distribution as the heat source. The results show that the power peaking factor FQ decreases below 1.65, the heat transfer area increases by 30.3% and the fuel peaking temperature decreases by 25 K. The optimized AM-informed design enjoys better neutronic and thermal-fluidic performance than those with regular geometry.
{"title":"Neutronic and Thermal-Fluidic Analyses for an Additive Manufactured Reactor with SiC Matrix and TRISO Particle Fuel","authors":"Wenbin Han, Jian Deng, Qi Lu, Chong Chen, Youyou Xu, Zhang Tao, Shanfang Huang","doi":"10.1115/1.4062119","DOIUrl":"https://doi.org/10.1115/1.4062119","url":null,"abstract":"\u0000 Additive manufacturing (AM) is a transformational digital manufacturing technology featured with rapidity, customizability, precision, and economy, which is fundamentally altering the way components are designed and manufactured. AM enables the freedom of design, and makes full use of complexity of geometry which \"comes for free\". Applying AM technology to nuclear industry can yield advanced reactor designs with function and structure matched for the best thermal, fluidic and mechanical performance. In this work, an AM-informed reactor core design with SiC matrix and TRISO particle fuel is proposed and analyzed. The core is an integrated 3D-printed SiC bulk with helical cruciform coolant channels, and the UO2-TRISO fuel particles are dispersed in the bulk. A multi-physics analysis framework for irregular geometry is developed to analyze and further optimize the reactor design. The TRISO particle positions are generated with discrete element method. The Reactor Monte Carlo code (RMC) and the CFD software STAR-CCM+ are used for the neutronic and thermal-fluidic analyses, respectively. RMC simulates the neutron transport to predict the effective multiplication factor and power distribution. STAR-CCM+ calculates the flow and heat transfer in coolant channels and heat conduction in solid matrix with the power distribution as the heat source. The results show that the power peaking factor FQ decreases below 1.65, the heat transfer area increases by 30.3% and the fuel peaking temperature decreases by 25 K. The optimized AM-informed design enjoys better neutronic and thermal-fluidic performance than those with regular geometry.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"23 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90714529","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}
Weng Wen, B. Ghidersa, W. Hering, J. Starflinger, R. Stieglitz
The use of heat pipes (HP) for the DEMO in-vessel plasma-facing components (PFCs) has been considered because of their high capacity to transport the heat from a heat source to a heat sink by means of the vaporization and condensation of the working fluid inside and their ability to enlarge the heat transfer area of the cooling circuit substantially. Recent engineering studies conducted in the framework of the EUROfusion work package Divertor (Wen et al, 2021) indicate that it is possible to design a heat pipe with a capillary limit above 6 kW using a composite capillary structure (wherein axial grooves cover the adiabatic zone and the condenser, and sintered porous material covers the evaporator). This power level would correspond to an applied heat flux of 20 MW/m2, rendering such a design interesting with respect to a divertor target concept. To validate the results of the initial engineering analysis, several experiments have been conducted to evaluate the actual performance of the proposed heat pipe concept. The present contribution presents the experiment’s results regarding the examination of the operating limits of two different designs for an evaporator: one featuring a plain porous structure, and one featuring ribs and channels.
由于热管能够通过内部工作流体的蒸发和冷凝将热量从热源输送到散热器，并且能够大大扩大冷却回路的传热面积，因此考虑了在DEMO容器内等离子体面组件(pfc)中使用热管(HP)。最近在EUROfusion工作包Divertor框架内进行的工程研究(Wen et al .， 2021)表明，可以使用复合毛细管结构(其中轴向槽覆盖绝热区和冷凝器，烧结多孔材料覆盖蒸发器)设计毛细管极限超过6 kW的热管。这一功率水平对应于20兆瓦/平方米的应用热流，使得这种设计在导流器目标概念方面很有趣。为了验证最初的工程分析结果，已经进行了几个实验来评估所提出的热管概念的实际性能。本文介绍了两种不同蒸发器设计的工作极限的实验结果:一种是平面多孔结构，另一种是肋和通道结构。
{"title":"Heat Pipe-Based DEMO Divertor Target Concept: High Heat Flux Performance Evaluation","authors":"Weng Wen, B. Ghidersa, W. Hering, J. Starflinger, R. Stieglitz","doi":"10.3390/jne4010021","DOIUrl":"https://doi.org/10.3390/jne4010021","url":null,"abstract":"The use of heat pipes (HP) for the DEMO in-vessel plasma-facing components (PFCs) has been considered because of their high capacity to transport the heat from a heat source to a heat sink by means of the vaporization and condensation of the working fluid inside and their ability to enlarge the heat transfer area of the cooling circuit substantially. Recent engineering studies conducted in the framework of the EUROfusion work package Divertor (Wen et al, 2021) indicate that it is possible to design a heat pipe with a capillary limit above 6 kW using a composite capillary structure (wherein axial grooves cover the adiabatic zone and the condenser, and sintered porous material covers the evaporator). This power level would correspond to an applied heat flux of 20 MW/m2, rendering such a design interesting with respect to a divertor target concept. To validate the results of the initial engineering analysis, several experiments have been conducted to evaluate the actual performance of the proposed heat pipe concept. The present contribution presents the experiment’s results regarding the examination of the operating limits of two different designs for an evaporator: one featuring a plain porous structure, and one featuring ribs and channels.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"15 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82648801","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}
Margot A. Hurlbert, Larissa Shasko, J. Condor, Dazawray Landrie-Parker
People’s affective response in relation to radiation is important in their risk perceptions of low-dose radiation (LDR), medical interventions involving LDR, and acceptance of nuclear power production. Risk perception studies generally relate to the health field of LDR or nuclear power. This study combines risk perceptions and acceptance of both. While acceptance by those with an understanding of radiation is demonstrated in focus groups, survey results disproved this correlation. Emotional response to the word radiation together with greater perceptions of risk to X-rays, were predictors of acceptance of nuclear power production.
{"title":"Radiation Workers and Risk Perceptions: Low Dose Radiation, Nuclear Power Production, and Small Modular Nuclear Reactors","authors":"Margot A. Hurlbert, Larissa Shasko, J. Condor, Dazawray Landrie-Parker","doi":"10.3390/jne4010020","DOIUrl":"https://doi.org/10.3390/jne4010020","url":null,"abstract":"People’s affective response in relation to radiation is important in their risk perceptions of low-dose radiation (LDR), medical interventions involving LDR, and acceptance of nuclear power production. Risk perception studies generally relate to the health field of LDR or nuclear power. This study combines risk perceptions and acceptance of both. While acceptance by those with an understanding of radiation is demonstrated in focus groups, survey results disproved this correlation. Emotional response to the word radiation together with greater perceptions of risk to X-rays, were predictors of acceptance of nuclear power production.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"70 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85118153","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}
Brian G. Frandsen, M. Febbraro, T. Ruland, Theodore W. Stephens, P. Hausladen, J. Manfredi, J. Bevins
Additive manufacturing techniques enable a wide range of possibilities for novel radiation detectors spanning simple to highly complex geometries, multi-material composites, and metamaterials that are either impossible or cost prohibitive to produce using conventional methods. The present work identifies a set of promising formulations of photocurable scintillator resins capable of neutron-gamma pulse shape discrimination (PSD) to support the additive manufacturing of fast neutron detectors. The development of these resins utilizes a step-by-step, trial-and-error approach to identify different monomer and cross-linker combinations that meet the requirements for 3D printing followed by a 2-level factorial parameter study to optimize the radiation detection performance, including light yield, PSD, optical clarity, and hardness. The formulations resulted in hard, clear, PSD-capable plastic scintillators that were cured solid within 10 s using 405 nm light. The best-performing scintillator produced a light yield 83% of EJ-276 and a PSD figure of merit equaling 1.28 at 450–550 keVee.
{"title":"Fast-, Light-Cured Scintillating Plastic for 3D-Printing Applications","authors":"Brian G. Frandsen, M. Febbraro, T. Ruland, Theodore W. Stephens, P. Hausladen, J. Manfredi, J. Bevins","doi":"10.3390/jne4010019","DOIUrl":"https://doi.org/10.3390/jne4010019","url":null,"abstract":"Additive manufacturing techniques enable a wide range of possibilities for novel radiation detectors spanning simple to highly complex geometries, multi-material composites, and metamaterials that are either impossible or cost prohibitive to produce using conventional methods. The present work identifies a set of promising formulations of photocurable scintillator resins capable of neutron-gamma pulse shape discrimination (PSD) to support the additive manufacturing of fast neutron detectors. The development of these resins utilizes a step-by-step, trial-and-error approach to identify different monomer and cross-linker combinations that meet the requirements for 3D printing followed by a 2-level factorial parameter study to optimize the radiation detection performance, including light yield, PSD, optical clarity, and hardness. The formulations resulted in hard, clear, PSD-capable plastic scintillators that were cured solid within 10 s using 405 nm light. The best-performing scintillator produced a light yield 83% of EJ-276 and a PSD figure of merit equaling 1.28 at 450–550 keVee.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"124 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75805038","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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