During the maintenance period of a future fusion reactor power plant, called DEMOnstration Power Plant (DEMO), remotely handled casks are required to confine and handle DEMO in-vessel components during their transportation between the reactor and the active maintenance facility. In order to limit the dispersion of activated dust, a Contamination Control Door (CCD) is designed to be placed at an interface between separable containments (e.g., vacuum vessels and casks) to inhibit the release of contamination at the interface between them. The remotely operated CCD—technically, a double lidded door system—consists of two separable doors (the cask door and port door) and three different locking mechanisms: (i) between the cask door and cask, (ii) between the cask door and port door and (iii) between the port door and port. The locking mechanisms are selected and assessed according to different criteria, and the structure of the CCD is optimized using an Abaqus Topology Optimization Module. Due to the elastic properties of the CCD, deflections will occur during the lifting procedure, which may lead to malfunctions of the CCD. A test rig is developed to investigate the performance of high-risk components in the CCD in the case of deflections and also malpositioning. Misalignment can be induced along three axes and three angles intentionally to test the single components and items. The aim is to identify a possible range of operating in the case of misalignments. It is expected that the proposed CCD design should be able to operate appropriately in the case of ±3 mm translational misalignments and ±1° rotational misalignments.
{"title":"Concept of Contamination Control Door for DEMO and Proof of Principle Design","authors":"Yan Wang, J. Oellerich, C. Baars, M. Mittwollen","doi":"10.3390/jne4010018","DOIUrl":"https://doi.org/10.3390/jne4010018","url":null,"abstract":"During the maintenance period of a future fusion reactor power plant, called DEMOnstration Power Plant (DEMO), remotely handled casks are required to confine and handle DEMO in-vessel components during their transportation between the reactor and the active maintenance facility. In order to limit the dispersion of activated dust, a Contamination Control Door (CCD) is designed to be placed at an interface between separable containments (e.g., vacuum vessels and casks) to inhibit the release of contamination at the interface between them. The remotely operated CCD—technically, a double lidded door system—consists of two separable doors (the cask door and port door) and three different locking mechanisms: (i) between the cask door and cask, (ii) between the cask door and port door and (iii) between the port door and port. The locking mechanisms are selected and assessed according to different criteria, and the structure of the CCD is optimized using an Abaqus Topology Optimization Module. Due to the elastic properties of the CCD, deflections will occur during the lifting procedure, which may lead to malfunctions of the CCD. A test rig is developed to investigate the performance of high-risk components in the CCD in the case of deflections and also malpositioning. Misalignment can be induced along three axes and three angles intentionally to test the single components and items. The aim is to identify a possible range of operating in the case of misalignments. It is expected that the proposed CCD design should be able to operate appropriately in the case of ±3 mm translational misalignments and ±1° rotational misalignments.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"36 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83616585","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}
X-ray radiography has proved to be essential in medical imaging and examination of material structures because it is non-invasive and generates images based on well-understood attenuation characteristics of materials. For radiographs of multiple overlapping materials, unraveling the individual attenuation contributions poses a problem that is commonly handled by either taking many radiographs at different object orientations for computed tomography or multiple images with different photon energies for Multiple Energy X-ray Absorptiometry (MEXA). Alternatively, to perform fast measurements, a novel algorithm has been developed to determine multi-material systems' density. The algorithm can be effectively applied to perform measurement using only one to four radiographs of the object. A case study has been presented for a layered cylindrical object that involved sensitivity studies on image noise, X-ray generator voltage fluctuations, layer thickness measurement perturbations, and X-ray generator photon energy distribution fluctuations using simulated radiographs and density calculations using actual radiographs. The results from the simulated and experimental results were found to agree with actual density values.
{"title":"A Novel Algorithm for Fast Measurement of Material Density in Symmetrical Objects Using X-Ray Radiography","authors":"V. Sinha, F. Strantz, Hyoung K. Lee","doi":"10.1115/1.4056972","DOIUrl":"https://doi.org/10.1115/1.4056972","url":null,"abstract":"\u0000 X-ray radiography has proved to be essential in medical imaging and examination of material structures because it is non-invasive and generates images based on well-understood attenuation characteristics of materials. For radiographs of multiple overlapping materials, unraveling the individual attenuation contributions poses a problem that is commonly handled by either taking many radiographs at different object orientations for computed tomography or multiple images with different photon energies for Multiple Energy X-ray Absorptiometry (MEXA). Alternatively, to perform fast measurements, a novel algorithm has been developed to determine multi-material systems' density. The algorithm can be effectively applied to perform measurement using only one to four radiographs of the object. A case study has been presented for a layered cylindrical object that involved sensitivity studies on image noise, X-ray generator voltage fluctuations, layer thickness measurement perturbations, and X-ray generator photon energy distribution fluctuations using simulated radiographs and density calculations using actual radiographs. The results from the simulated and experimental results were found to agree with actual density values.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"17 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83795140","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}
Tommaso Del Moro, P. Lorusso, F. Giannetti, M. Tarantino, M. Caramello, D. Mazzi, M. Constantin
Among the envisaged experimental infrastructures supporting ALFRED reactor development, FALCON consortium identified ATHENA as one of the facilities to address the pool thermal-hydraulic challenges and demonstrate the feasibility of the revised ALFRED configuration, along with the thermal-hydraulic performances of its main components. ATHENA is a large pool-type lead-cooled multi-purpose experimental facility featuring a large size vessel (3.2 m diameter, 10 m in height), conceived to host almost 800 tons of lead to test ALFRED relevant scaled components. The test section to be installed in the main vessel includes an electrically heated core simulator, made of 7 FAs, which delivers to the primary coolant a nominal thermal power of 2210 kW, a main coolant pump for lead circulation and a counter-current shell and tubes main heat exchanger, which tube bundle is fed by pressurized water by a dedicated secondary circuit. A preliminary analytical calculation has been performed to identify the most suitable configuration for the ATHENA heat exchanger, as well as to achieve a preliminary sizing of the component. The final layout foresees a bundle of 91 bayonet tubes, arranged in a cylindrical geometry, having an active length of 3000 mm and adopting an external double wall tube with a pressurized helium gap. A numerical model of the heat exchanger has been developed using the thermal-hydraulic system code RELAP5/Mod3.3, and a numerical sensitivity analysis on the geometrical and operating parameters has been carried out to verify the performances of the component.
{"title":"Athena Main Heat Exchanger Conceptual Design and Thermal-Hydraulic Assessment with RELAP5 Code","authors":"Tommaso Del Moro, P. Lorusso, F. Giannetti, M. Tarantino, M. Caramello, D. Mazzi, M. Constantin","doi":"10.1115/1.4056973","DOIUrl":"https://doi.org/10.1115/1.4056973","url":null,"abstract":"\u0000 Among the envisaged experimental infrastructures supporting ALFRED reactor development, FALCON consortium identified ATHENA as one of the facilities to address the pool thermal-hydraulic challenges and demonstrate the feasibility of the revised ALFRED configuration, along with the thermal-hydraulic performances of its main components. ATHENA is a large pool-type lead-cooled multi-purpose experimental facility featuring a large size vessel (3.2 m diameter, 10 m in height), conceived to host almost 800 tons of lead to test ALFRED relevant scaled components. The test section to be installed in the main vessel includes an electrically heated core simulator, made of 7 FAs, which delivers to the primary coolant a nominal thermal power of 2210 kW, a main coolant pump for lead circulation and a counter-current shell and tubes main heat exchanger, which tube bundle is fed by pressurized water by a dedicated secondary circuit. A preliminary analytical calculation has been performed to identify the most suitable configuration for the ATHENA heat exchanger, as well as to achieve a preliminary sizing of the component. The final layout foresees a bundle of 91 bayonet tubes, arranged in a cylindrical geometry, having an active length of 3000 mm and adopting an external double wall tube with a pressurized helium gap. A numerical model of the heat exchanger has been developed using the thermal-hydraulic system code RELAP5/Mod3.3, and a numerical sensitivity analysis on the geometrical and operating parameters has been carried out to verify the performances of the component.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"27 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87044699","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}
Haiyan Yu, Guifeng Zhu, Y. Zou, Rui Yan, Yafen Liu, Xuzhong Kang, Ye Dai
The SM-MSR (small modular molten salt reactor) has a good prospect for development with regards to combining the superiority of the molten salt reactor and modularization technologies, showing the advantages of safety, reliability, low economic cost and flexibility of site selection. However, because its internal structural parts are not easily replaced, and the outer shielding structure is limited, the lifespan of the reactor vessel and its in-reactor shielding design needs to be addressed. In order to find an optimal shielding model with both high fuel efficiency and strong radiation shielding capability, five different design schemes were proposed in this work, which varied in thickness and boron concentration in inner-shielding materials. The neutron/gamma flux and DPA (displacements per atom)/helium production rates were evaluated to obtain an appropriate scheme. Several beneficial results were obtained. Considering the above factors and the actual manufacturing process, 20 cm-thick boron graphite with a 5 wt% Boron-10 concentration combined with a 1 cm-thick Hastelloy barrel was chosen as the in-reactor shielding structure. Outside the reactor, the neutron flux was reduced to 8.33 × 1010 cm−2 s−1, and the gamma flux was decreased to 1.13 × 1011 cm−2 s−1. The vessel/barrel material could maintain a lifespan of more than 10 years, while the burnup depth was 6.25% lower than that of a model without inner-shielding. The conclusions of this research can provide important references for the shielding design and parameter selections of small molten salt reactors in the future.
{"title":"Neutron/Gamma Radial Shielding Design of Main Vessel in a Small Modular Molten Salt Reactor","authors":"Haiyan Yu, Guifeng Zhu, Y. Zou, Rui Yan, Yafen Liu, Xuzhong Kang, Ye Dai","doi":"10.3390/jne4010017","DOIUrl":"https://doi.org/10.3390/jne4010017","url":null,"abstract":"The SM-MSR (small modular molten salt reactor) has a good prospect for development with regards to combining the superiority of the molten salt reactor and modularization technologies, showing the advantages of safety, reliability, low economic cost and flexibility of site selection. However, because its internal structural parts are not easily replaced, and the outer shielding structure is limited, the lifespan of the reactor vessel and its in-reactor shielding design needs to be addressed. In order to find an optimal shielding model with both high fuel efficiency and strong radiation shielding capability, five different design schemes were proposed in this work, which varied in thickness and boron concentration in inner-shielding materials. The neutron/gamma flux and DPA (displacements per atom)/helium production rates were evaluated to obtain an appropriate scheme. Several beneficial results were obtained. Considering the above factors and the actual manufacturing process, 20 cm-thick boron graphite with a 5 wt% Boron-10 concentration combined with a 1 cm-thick Hastelloy barrel was chosen as the in-reactor shielding structure. Outside the reactor, the neutron flux was reduced to 8.33 × 1010 cm−2 s−1, and the gamma flux was decreased to 1.13 × 1011 cm−2 s−1. The vessel/barrel material could maintain a lifespan of more than 10 years, while the burnup depth was 6.25% lower than that of a model without inner-shielding. The conclusions of this research can provide important references for the shielding design and parameter selections of small molten salt reactors in the future.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"424 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84935332","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}
Plasma-facing materials (PFMs) in the ITER divertor will be exposed to severe conditions, including exposure to transient heat loads from edge-localized modes (ELMs) and to plasma particles and neutrons. Tungsten is the material chosen as PFM for the ITER divertor. In previous tests, bubble formation in ITER-grade tungsten was detected when exposed to fusion relevant conditions. For this study, ITER-grade tungsten was exposed to simultaneous ELM-like transient heat loads and D/He (6%) plasma in the linear plasma device PSI-2. Bubble formation was then investigated via SEM micrographs and FIB cuts. It was found that for exposure to 100.000 laser pulses of 0.6 GWm−2 absorbed power density (Pabs), only small bubbles in the nanometer range were formed close to the surface. After increasing Pabs to 0.8 and 1.0 GWm−2, the size of the bubbles went up to about 1 µm in size and were deeper below the surface. Increasing the plasma fluence had an even larger effect, more than doubling bubble density and increasing bubble size to up to 2 µm in diameter. When using deuterium-only plasma, the samples showed no bubble formation and reduced cracking, showing such bubble formation is caused by exposure to helium plasma.
{"title":"Bubble Formation in ITER-Grade Tungsten after Exposure to Stationary D/He Plasma and ELM-like Thermal Shocks","authors":"M. Gago, A. Kreter, B. Unterberg, M. Wirtz","doi":"10.3390/jne4010016","DOIUrl":"https://doi.org/10.3390/jne4010016","url":null,"abstract":"Plasma-facing materials (PFMs) in the ITER divertor will be exposed to severe conditions, including exposure to transient heat loads from edge-localized modes (ELMs) and to plasma particles and neutrons. Tungsten is the material chosen as PFM for the ITER divertor. In previous tests, bubble formation in ITER-grade tungsten was detected when exposed to fusion relevant conditions. For this study, ITER-grade tungsten was exposed to simultaneous ELM-like transient heat loads and D/He (6%) plasma in the linear plasma device PSI-2. Bubble formation was then investigated via SEM micrographs and FIB cuts. It was found that for exposure to 100.000 laser pulses of 0.6 GWm−2 absorbed power density (Pabs), only small bubbles in the nanometer range were formed close to the surface. After increasing Pabs to 0.8 and 1.0 GWm−2, the size of the bubbles went up to about 1 µm in size and were deeper below the surface. Increasing the plasma fluence had an even larger effect, more than doubling bubble density and increasing bubble size to up to 2 µm in diameter. When using deuterium-only plasma, the samples showed no bubble formation and reduced cracking, showing such bubble formation is caused by exposure to helium plasma.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"24 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75382798","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}
Ben Lindley, Francisco Álvarez Velarde, Una Baker, J. Bodi, P. Cosgrove, Alan Charles, C. Fiorina, E. Fridman, J. Křepel, J. Lavarenne, K. Mikityuk, E. Nikitin, A. Ponomarev, S. Radman, E. Shwageraus, B. Tollit
The objective of this paper is to quantify the coupling effect on the power distribution of sodium-cooled fast reactors (SFRs), specifically the European SFR. Calculations are performed with several state-of-the-art reactor physics and Multiphysics codes (TRACE/PARCS, DYN3D, WIMS, COUNTHER and GeN-Foam) to build confidence in the methodologies and validity of results. Standalone neutronics calculations were generally in excellent agreement with a reference Monte Carlo-calculated power distribution (from Serpent). Next, the impact of coolant density and fuel temperature Doppler feedback was calculated. Reactivity coefficients for perturbations in the inlet temperature, flow rate and core power were shown to be negative with values of around -0.5 pcm/°C, -0.3 pcm/°C and -3.5 pcm/% respectively. Fuel temperature and coolant density feedback was found to introduce a roughly -1%/+1% in/out power tilt across the core. Calculations were then extended to axial expansion for cases where fuel is linked and unlinked to the clad. Core calculations are in good agreement with each other. The impact of differential fuel expansion is found to be larger for fuel both linked and unlinked to the clad, with the in/out power tilt increasing to around -4%/+2%. Thus, while broadly confirming the known result that standalone physics calculations give good results, the expansion coupling effect is perhaps more than anticipated a priori. These results provide a useful benchmark for the further development of Multiphysics codes and methodologies in support of advanced reactor calculations.
{"title":"Impact of Thermal-hydraulic Feedback and Differential Thermal Expansion On European Sfr Core Power Distribution","authors":"Ben Lindley, Francisco Álvarez Velarde, Una Baker, J. Bodi, P. Cosgrove, Alan Charles, C. Fiorina, E. Fridman, J. Křepel, J. Lavarenne, K. Mikityuk, E. Nikitin, A. Ponomarev, S. Radman, E. Shwageraus, B. Tollit","doi":"10.1115/1.4056930","DOIUrl":"https://doi.org/10.1115/1.4056930","url":null,"abstract":"\u0000 The objective of this paper is to quantify the coupling effect on the power distribution of sodium-cooled fast reactors (SFRs), specifically the European SFR. Calculations are performed with several state-of-the-art reactor physics and Multiphysics codes (TRACE/PARCS, DYN3D, WIMS, COUNTHER and GeN-Foam) to build confidence in the methodologies and validity of results. Standalone neutronics calculations were generally in excellent agreement with a reference Monte Carlo-calculated power distribution (from Serpent). Next, the impact of coolant density and fuel temperature Doppler feedback was calculated. Reactivity coefficients for perturbations in the inlet temperature, flow rate and core power were shown to be negative with values of around -0.5 pcm/°C, -0.3 pcm/°C and -3.5 pcm/% respectively. Fuel temperature and coolant density feedback was found to introduce a roughly -1%/+1% in/out power tilt across the core. Calculations were then extended to axial expansion for cases where fuel is linked and unlinked to the clad. Core calculations are in good agreement with each other. The impact of differential fuel expansion is found to be larger for fuel both linked and unlinked to the clad, with the in/out power tilt increasing to around -4%/+2%. Thus, while broadly confirming the known result that standalone physics calculations give good results, the expansion coupling effect is perhaps more than anticipated a priori. These results provide a useful benchmark for the further development of Multiphysics codes and methodologies in support of advanced reactor calculations.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"63 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84082282","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}
Olugbenga O Noah, Johan . F Slabber, Josua P Meyer
Abstract The development of an accident tolerant nuclear fuel for water-cooled reactors would redefined the status of these reactors from traditional active safety to passive safety systems. As a possible solution toward enhancing the safety of light-water reactors (LWRs), loose-coated particles of enriched uranium dioxide (UO2) fuel with the ability to retain gaseous and metallic fission products in the case of a loss of cooling event can be introduced inside Silicon-Carbide cladding tubes of the fuel assembly (see Figs. 1(a) and 1(b)). These coated particles are treated as a bed from where heat is transferred to the cladding tube and the helium gas movement is due to natural convection. A slender geometrical model with tube-to-particle diameter ratio N = 2.503 and porosity ε = 0.546 mimicking the proposed nuclear fuel in the cladding was numerically simulated. This study is to investigate the heat transfer characteristics and flow distribution under buoyancy driven force expected in the cladding tube of the proposed nuclear fuel using a commercial code. Random packing of the particles is achieved by discrete element method (DEM) simulation with the aid of starccm+. The temperature contour and velocity vector plots obtained can be said to be good illustration of anticipated heat transfer and transport phenomenon to occur in the proposed fuel design. Simulated results for particle-to-fluid heat transfer coefficient, Nusselt number, and Rayleigh number which are of prime importance when analyzing natural convection heat transfer performance in fixed bed reactors were validated. Results from this work show close agreement with results obtained in established numerical and experimental works.
{"title":"Introducing Passive Nuclear Safety in Water-Cooled Reactors: Numerical Simulation and Validation of Natural Convection Heat Transfer and Transport in Packed Beds of Heated Microspheres","authors":"Olugbenga O Noah, Johan . F Slabber, Josua P Meyer","doi":"10.1115/1.4056239","DOIUrl":"https://doi.org/10.1115/1.4056239","url":null,"abstract":"Abstract The development of an accident tolerant nuclear fuel for water-cooled reactors would redefined the status of these reactors from traditional active safety to passive safety systems. As a possible solution toward enhancing the safety of light-water reactors (LWRs), loose-coated particles of enriched uranium dioxide (UO2) fuel with the ability to retain gaseous and metallic fission products in the case of a loss of cooling event can be introduced inside Silicon-Carbide cladding tubes of the fuel assembly (see Figs. 1(a) and 1(b)). These coated particles are treated as a bed from where heat is transferred to the cladding tube and the helium gas movement is due to natural convection. A slender geometrical model with tube-to-particle diameter ratio N = 2.503 and porosity ε = 0.546 mimicking the proposed nuclear fuel in the cladding was numerically simulated. This study is to investigate the heat transfer characteristics and flow distribution under buoyancy driven force expected in the cladding tube of the proposed nuclear fuel using a commercial code. Random packing of the particles is achieved by discrete element method (DEM) simulation with the aid of starccm+. The temperature contour and velocity vector plots obtained can be said to be good illustration of anticipated heat transfer and transport phenomenon to occur in the proposed fuel design. Simulated results for particle-to-fluid heat transfer coefficient, Nusselt number, and Rayleigh number which are of prime importance when analyzing natural convection heat transfer performance in fixed bed reactors were validated. Results from this work show close agreement with results obtained in established numerical and experimental works.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"91 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"136175782","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 In the design study of an advanced sodium-cooled fast reactor (advanced-SFR) in Japan Atomic Energy Agency (JAEA), the use of a specific fuel assembly (FA) with an inner duct structure called fuel assembly with an inner duct structure (FAIDUS) has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the coolant temperature distribution to confirm feasibility of FAIDUS. In JAEA, an in-house subchannel analysis code named thermal-hydraulic analysis of asymmetrical flow in reactor elements (ASFRE) has been developed as a FA design tool. For the typical FAs, the numerical results of ASFRE had been validated by comparisons with experimental data, in the previous study. As for the FAIDUS, confirmation of validity of the numerical results by ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mockup experiment, yet. In this paper, therefore, the code-to-code comparisons with numerical results of ASFRE and those of an in-house computational fluid dynamics (CFD) code named single-phase thermal-hydraulic analysis in an irregular rod array layout (SPIRAL) were applied to make further discussion on applicability of ASFRE to the thermal hydraulics analysis in FAIDUS. Thermal hydraulic analyses of a typical FA and FAIDUS at high and low flowrate conditions were conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism on appearance of the specific temperature distributions between the numerical results by ASFRE and those by SPIRAL. In addition, the necessity of modification on the empirical constants in numerical model of ASFRE to improve the predictive accuracy was indicated.
{"title":"Investigation on Applicability of Subchannel Analysis Code ASFRE to Thermal Hydraulics Analysis in Fuel Assembly With Inner Duct Structure of Sodium Cooled Fast Reactor","authors":"Norihiro Kikuchi, Yasutomo Imai, Ryuji Yoshikawa, Norihiro Doda, Masaaki Tanaka","doi":"10.1115/1.4056463","DOIUrl":"https://doi.org/10.1115/1.4056463","url":null,"abstract":"Abstract In the design study of an advanced sodium-cooled fast reactor (advanced-SFR) in Japan Atomic Energy Agency (JAEA), the use of a specific fuel assembly (FA) with an inner duct structure called fuel assembly with an inner duct structure (FAIDUS) has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the coolant temperature distribution to confirm feasibility of FAIDUS. In JAEA, an in-house subchannel analysis code named thermal-hydraulic analysis of asymmetrical flow in reactor elements (ASFRE) has been developed as a FA design tool. For the typical FAs, the numerical results of ASFRE had been validated by comparisons with experimental data, in the previous study. As for the FAIDUS, confirmation of validity of the numerical results by ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mockup experiment, yet. In this paper, therefore, the code-to-code comparisons with numerical results of ASFRE and those of an in-house computational fluid dynamics (CFD) code named single-phase thermal-hydraulic analysis in an irregular rod array layout (SPIRAL) were applied to make further discussion on applicability of ASFRE to the thermal hydraulics analysis in FAIDUS. Thermal hydraulic analyses of a typical FA and FAIDUS at high and low flowrate conditions were conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism on appearance of the specific temperature distributions between the numerical results by ASFRE and those by SPIRAL. In addition, the necessity of modification on the empirical constants in numerical model of ASFRE to improve the predictive accuracy was indicated.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"128 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"136175636","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}
In this work we report the characterization of thin metallic coatings of interest for nuclear fusion technology through the ns double-pulse LIBS technique. The coatings, composed of a tungsten (W) or tungsten-tantalum (W-Ta) mixture were enriched with deuterium (D), to simulate plasma-facing materials (PFMs) or components (PFCs) of the next generation devices contaminated with nuclear fuel in the divertor area of the vacuum vessel (VV), with special attention to ITER, whose divertor will be made of W. The double pulse LIBS technique allowed for the detection of D and Ta at low concentrations, with a single laser shot and an average ablation rate of about 110 nm. The calibration free (CF-LIBS) procedure provided a semi-quantitative estimation of the retained deuterium in the coatings, without the need of reference samples. The presented results demonstrate that LIBS is an eligible diagnostic tool to characterize PFCs with high sensitivity and accuracy, being minimally destructive on the samples, without PFCs manipulation. The CF-LIBS procedure can be used for the search for any other materials in the VV without any preliminary reference samples.
{"title":"Double Pulse LIBS Analysis of Metallic Coatings of Fusionistic Interest: Depth Profiling and Semi-Quantitative Elemental Composition by Applying the Calibration Free Technique","authors":"S. Almaviva, F. Colao, I. Menicucci, M. Pistilli","doi":"10.3390/jne4010015","DOIUrl":"https://doi.org/10.3390/jne4010015","url":null,"abstract":"In this work we report the characterization of thin metallic coatings of interest for nuclear fusion technology through the ns double-pulse LIBS technique. The coatings, composed of a tungsten (W) or tungsten-tantalum (W-Ta) mixture were enriched with deuterium (D), to simulate plasma-facing materials (PFMs) or components (PFCs) of the next generation devices contaminated with nuclear fuel in the divertor area of the vacuum vessel (VV), with special attention to ITER, whose divertor will be made of W. The double pulse LIBS technique allowed for the detection of D and Ta at low concentrations, with a single laser shot and an average ablation rate of about 110 nm. The calibration free (CF-LIBS) procedure provided a semi-quantitative estimation of the retained deuterium in the coatings, without the need of reference samples. The presented results demonstrate that LIBS is an eligible diagnostic tool to characterize PFCs with high sensitivity and accuracy, being minimally destructive on the samples, without PFCs manipulation. The CF-LIBS procedure can be used for the search for any other materials in the VV without any preliminary reference samples.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"165 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-02-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78567268","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}
Gangadharayya B. Hiremath, M. Hosamani, V. P. Singh, N. Ayachit, Nagappa M. Badiger
Inorganic scintillators are used in extreme environments is essential in applications such as nuclear fusion, marine and space exploration, etc. Radiation resistance is requirement for scintillators to be used in above mentioned applications. It is important to calculate the gamma buildup factors for inorganic scintillators to understand its radiation resistance behaviour. The gamma-ray and neutron interaction parameters were calculated for inorganic scintillators such as BaF2, BGO, CdWO4, CaF2(Eu), CaWO4, CsI, CsI(Na), CsI(Tl), Gd2O2S, LaBr3(Ce), LaCl3(Ce), PbWO4, LSO(Ce), LYSO(Ce), NaI(Tl), YAG(Ce), ZnS(Ag), ZnWO4. The gamma ray interaction parameters such as mass attenuation coefficient, half value layer, tenth value layer, effective atomic number, equivalent atomic number, exposure and energy absorption buildup factor were computed using Phy-X/PSD software. The neutron interaction parameters such as fast neutron removal crosssection, and mass attenuation factor for thermal and fast neutron were computed using Phy-X/PSD and NGCal software respectively. The present calculated mass attenuation coefficient values are compared with geant4 data available in the literature. The higher values of Zeff were found for PbWO4 upto 0.1 MeV and above 0.1 MeV, CsI(Tl) has higher Zeff values. However, the Zeq values were found to be higher for PbWO4 in the whole energy range. The mass attenuation factor for thermal and fast neutron was found that Gd2O2S has highest value and BGO has lowest value. The present investigation indicates that the inorganic scintillator PbWO4 has an excellent gamma-ray sensing response.
{"title":"Theoretical Investigation of the Gamma and Neutron Interaction Parameters of Some Inorganic Scintillators Using Phy-X/PSD and NGCal Software","authors":"Gangadharayya B. Hiremath, M. Hosamani, V. P. Singh, N. Ayachit, Nagappa M. Badiger","doi":"10.1115/1.4056835","DOIUrl":"https://doi.org/10.1115/1.4056835","url":null,"abstract":"\u0000 Inorganic scintillators are used in extreme environments is essential in applications such as nuclear fusion, marine and space exploration, etc. Radiation resistance is requirement for scintillators to be used in above mentioned applications. It is important to calculate the gamma buildup factors for inorganic scintillators to understand its radiation resistance behaviour. The gamma-ray and neutron interaction parameters were calculated for inorganic scintillators such as BaF2, BGO, CdWO4, CaF2(Eu), CaWO4, CsI, CsI(Na), CsI(Tl), Gd2O2S, LaBr3(Ce), LaCl3(Ce), PbWO4, LSO(Ce), LYSO(Ce), NaI(Tl), YAG(Ce), ZnS(Ag), ZnWO4. The gamma ray interaction parameters such as mass attenuation coefficient, half value layer, tenth value layer, effective atomic number, equivalent atomic number, exposure and energy absorption buildup factor were computed using Phy-X/PSD software. The neutron interaction parameters such as fast neutron removal crosssection, and mass attenuation factor for thermal and fast neutron were computed using Phy-X/PSD and NGCal software respectively. The present calculated mass attenuation coefficient values are compared with geant4 data available in the literature. The higher values of Zeff were found for PbWO4 upto 0.1 MeV and above 0.1 MeV, CsI(Tl) has higher Zeff values. However, the Zeq values were found to be higher for PbWO4 in the whole energy range. The mass attenuation factor for thermal and fast neutron was found that Gd2O2S has highest value and BGO has lowest value. The present investigation indicates that the inorganic scintillator PbWO4 has an excellent gamma-ray sensing response.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"28 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-02-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86829016","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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