Koji Morita, W. Liu, T. Arima, Yuji Arita, Isamu Sato, H. Matsuura, Y. Sekio, H. Sagara, Masatoshi Kawashima
Following the Fukushima Nuclear Power Plant accident in 2011, it has become increasingly important for reactor safety designs to consider measures that can prevent the occurrence of severe accidents. This report proposes a novel subassembly-type passive reactor shutdown device that expands the diversity and robustness of core disruptive accident prevention strategies for sodium-cooled fast reactors. The developed device contains pins with a fuel material that is in the solid state during normal operation but melts into a liquid when the temperature exceeds a certain value (i.e., during a potential accident). When an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident occurs, the device can passively provide significant negative reactivity by rapidly transferring liquefied device fuel into the lower plenum region of the pins via gravitation alone. The reactors containing some of the proposed devices in place of original fuel subassemblies become subcritical before the driver fuels are damaged, even if ULOF or UTOP transient events occur. The present study evaluates candidate materials for device fuels (e.g., metallic alloy, chloride), optimal device pin structures for liquefied fuel relocation, and nuclear and thermal-hydraulic characteristics of the device-loaded core under accident conditions to demonstrate the engineering applicability of the proposed device. This report discusses preliminary results regarding the nuclear requirements for inducing negative reactivity to achieve reactor shutdown under the expected device conditions during an accident.
{"title":"Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Preliminary Study","authors":"Koji Morita, W. Liu, T. Arima, Yuji Arita, Isamu Sato, H. Matsuura, Y. Sekio, H. Sagara, Masatoshi Kawashima","doi":"10.1115/1.4056834","DOIUrl":"https://doi.org/10.1115/1.4056834","url":null,"abstract":"\u0000 Following the Fukushima Nuclear Power Plant accident in 2011, it has become increasingly important for reactor safety designs to consider measures that can prevent the occurrence of severe accidents. This report proposes a novel subassembly-type passive reactor shutdown device that expands the diversity and robustness of core disruptive accident prevention strategies for sodium-cooled fast reactors. The developed device contains pins with a fuel material that is in the solid state during normal operation but melts into a liquid when the temperature exceeds a certain value (i.e., during a potential accident). When an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident occurs, the device can passively provide significant negative reactivity by rapidly transferring liquefied device fuel into the lower plenum region of the pins via gravitation alone. The reactors containing some of the proposed devices in place of original fuel subassemblies become subcritical before the driver fuels are damaged, even if ULOF or UTOP transient events occur. The present study evaluates candidate materials for device fuels (e.g., metallic alloy, chloride), optimal device pin structures for liquefied fuel relocation, and nuclear and thermal-hydraulic characteristics of the device-loaded core under accident conditions to demonstrate the engineering applicability of the proposed device. This report discusses preliminary results regarding the nuclear requirements for inducing negative reactivity to achieve reactor shutdown under the expected device conditions during an accident.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"114 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-02-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88141892","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}
The outcome of the electromagnetic (EM) analyses carried out during the DEMO pre-conceptual phase demonstrated that EM loads are relevant for the structural assessment of the breeding blanket (BB) and, in particular, for the definition of the boundary conditions at the attachment system with the vacuum vessel. However, within the scope of the previous campaign, the results obtained using simplified models only give a rough estimation of the EM loads inside the BB structure. This kind of data has been considered suitable for a preliminary assessment of the BB segments, but it is not considered representative as input for structural analysis in which a detailed BB internal structure (that considers cooling channels, thin plates, etc.) is analyzed. Indeed, mesh dimensions and computational time usually limit EM models that simulate a whole DEMO sector. In many cases, these constraints lead to a strong homogenization of the BB structure, not allowing the calculation of the EM loads on the internal structure with high precision. To overcome such limitations, an EM sub-modeling procedure was investigated using ANSYS EMAG. The sub-modeling feasibility is studied using the rigid boundary condition method. This method consists of running a global “coarse” mesh, including all the conducting structures that can have some impact on the component under investigation and inputting the obtained results on the detailed sub-model of the structure of interest as time-varying boundary conditions. The procedure was tested on the BB internal structure, taking as reference a DEMO 2017 baseline sector and the helium cooled pebble bed (HCPB) concept with its complex internal structure made by pins. The obtained results show that the method is also reliable in the presence of non-linear magnetic behaviour. The methodology is proposed for application in future BB system assessments.
{"title":"Investigation of Electromagnetic Sub-Modeling Procedure for the Breeding Blanket System","authors":"I. Maione, M. Roccella, Flavio Lucca","doi":"10.3390/jne4010013","DOIUrl":"https://doi.org/10.3390/jne4010013","url":null,"abstract":"The outcome of the electromagnetic (EM) analyses carried out during the DEMO pre-conceptual phase demonstrated that EM loads are relevant for the structural assessment of the breeding blanket (BB) and, in particular, for the definition of the boundary conditions at the attachment system with the vacuum vessel. However, within the scope of the previous campaign, the results obtained using simplified models only give a rough estimation of the EM loads inside the BB structure. This kind of data has been considered suitable for a preliminary assessment of the BB segments, but it is not considered representative as input for structural analysis in which a detailed BB internal structure (that considers cooling channels, thin plates, etc.) is analyzed. Indeed, mesh dimensions and computational time usually limit EM models that simulate a whole DEMO sector. In many cases, these constraints lead to a strong homogenization of the BB structure, not allowing the calculation of the EM loads on the internal structure with high precision. To overcome such limitations, an EM sub-modeling procedure was investigated using ANSYS EMAG. The sub-modeling feasibility is studied using the rigid boundary condition method. This method consists of running a global “coarse” mesh, including all the conducting structures that can have some impact on the component under investigation and inputting the obtained results on the detailed sub-model of the structure of interest as time-varying boundary conditions. The procedure was tested on the BB internal structure, taking as reference a DEMO 2017 baseline sector and the helium cooled pebble bed (HCPB) concept with its complex internal structure made by pins. The obtained results show that the method is also reliable in the presence of non-linear magnetic behaviour. The methodology is proposed for application in future BB system assessments.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"41 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90707415","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}
The mechanical consequences of core disruptive accidents (CDAs) are a major safety concern in sodium-cooled fast reactors. Once core disruption occurs, liquefied core materials rapidly disperse vertically and radially. The dispersed materials penetrate the pin bundles and control rod guide tubes (CRGTs) before freezing at the edge of the penetration zone as heat is transferred to surrounding structures. Such freezing phenomena can suppress the negative reactivity feedback of fuel dispersion. The discharge of core materials can be impeded, resulting in a molten core pool formation when tight blockages occur inside CRGTs due to frozen material. Accordingly, freezing phenomena of core materials play a key role in governing the mechanical consequences of a CDA. To validate a freezing model implemented in our CDA analysis code, ASTERIA-SFR, a preliminary simulation of the THEFIS RUN#1 test, was performed. The calculation results show that freezing on the structural wall and crust formation were key phenomena affecting the penetration behavior, and the structural heat transfer is an important parameter. A remarkable reduction of the heat transfer coefficient was required to reproduce the penetration length observed in the experiment. This suggests that the momentum exchange and flow regime at the leading edge as well as heat transfer should be well modeled to predict the freezing phenomena in rapidly evolving CDAs.
{"title":"THEFIS Test Simulation to Validate a Freezing Model of ASTERIA-SFR Core Disruptive Accident Analysis Code","authors":"T. Ishizu, Hiroki Sonoda, S. Fujita","doi":"10.3390/jne4010012","DOIUrl":"https://doi.org/10.3390/jne4010012","url":null,"abstract":"The mechanical consequences of core disruptive accidents (CDAs) are a major safety concern in sodium-cooled fast reactors. Once core disruption occurs, liquefied core materials rapidly disperse vertically and radially. The dispersed materials penetrate the pin bundles and control rod guide tubes (CRGTs) before freezing at the edge of the penetration zone as heat is transferred to surrounding structures. Such freezing phenomena can suppress the negative reactivity feedback of fuel dispersion. The discharge of core materials can be impeded, resulting in a molten core pool formation when tight blockages occur inside CRGTs due to frozen material. Accordingly, freezing phenomena of core materials play a key role in governing the mechanical consequences of a CDA. To validate a freezing model implemented in our CDA analysis code, ASTERIA-SFR, a preliminary simulation of the THEFIS RUN#1 test, was performed. The calculation results show that freezing on the structural wall and crust formation were key phenomena affecting the penetration behavior, and the structural heat transfer is an important parameter. A remarkable reduction of the heat transfer coefficient was required to reproduce the penetration length observed in the experiment. This suggests that the momentum exchange and flow regime at the leading edge as well as heat transfer should be well modeled to predict the freezing phenomena in rapidly evolving CDAs.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"32 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76901405","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}
High-quality academic publishing is built on rigorous peer review [...]
高质量的学术出版建立在严格的同行评审的基础上[…]
{"title":"Acknowledgment to the Reviewers of JNE in 2022","authors":"","doi":"10.3390/jne4010011","DOIUrl":"https://doi.org/10.3390/jne4010011","url":null,"abstract":"High-quality academic publishing is built on rigorous peer review [...]","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89859445","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}
D. Nagy, S. Zoletnik, M. Otte, M. Vécsei, M. Krychowiak, R. König, D. Dunai, G. Anda, S. Hegedűs, B. Csillag, I. Katona
On a Wendelstein 7-X (W7-X), an alkali metal beam (AMB) diagnostic system was installed in order to measure the plasma edge electron density profiles and turbulence transport. A sodium beam was injected in the plasma, and the light emission was observed by an optical system. During the last operation phase, OP1.2b campaign trial spectral measurements were performed with a dedicated optical branch. The results showed the emergence of potential CX lines in the light spectra during sodium injection. The lines were identified as Carbon III, which were the dominant lines observed by other diagnostics at the edge plasma. Based on these results, an additional dedicated optical system was developed and installed in 2021 for the upcoming operational phase, OP2. The optics were designed for multiple purposes: spectral measurements for the AMB system and for a He/Ne gas jet. The system was designed to allow implementation of further diagnostics on this port later (e.g., coherence imaging system). The details of the implementation of the design requirements and the main challenges of the manufacturing process and installation are discussed in this paper.
{"title":"Development of the W7-X Alkali Metal Beam Diagnostic Observation System for OP2","authors":"D. Nagy, S. Zoletnik, M. Otte, M. Vécsei, M. Krychowiak, R. König, D. Dunai, G. Anda, S. Hegedűs, B. Csillag, I. Katona","doi":"10.3390/jne4010010","DOIUrl":"https://doi.org/10.3390/jne4010010","url":null,"abstract":"On a Wendelstein 7-X (W7-X), an alkali metal beam (AMB) diagnostic system was installed in order to measure the plasma edge electron density profiles and turbulence transport. A sodium beam was injected in the plasma, and the light emission was observed by an optical system. During the last operation phase, OP1.2b campaign trial spectral measurements were performed with a dedicated optical branch. The results showed the emergence of potential CX lines in the light spectra during sodium injection. The lines were identified as Carbon III, which were the dominant lines observed by other diagnostics at the edge plasma. Based on these results, an additional dedicated optical system was developed and installed in 2021 for the upcoming operational phase, OP2. The optics were designed for multiple purposes: spectral measurements for the AMB system and for a He/Ne gas jet. The system was designed to allow implementation of further diagnostics on this port later (e.g., coherence imaging system). The details of the implementation of the design requirements and the main challenges of the manufacturing process and installation are discussed in this paper.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"26 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-01-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83822065","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}
D. Zhan, Jihang Li, Dongwei Wang, Hui-shu Zhang, Guoxing Qiu, Yongkun Yang
In this study, we present the effects of 0.004~0.098 wt% Zr and thermo-mechanical processing (TMP) on the microstructure and mechanical properties of the China RAFM steel, CLAM, as a feasibility study for improving mechanical properties. The inclusions in ingots were characterized using optical microscope (OM) and scanning electron microscope (SEM), which could be classified as fine simple particles and large complex particles. The complexity of the alloy’s inclusion composition increases with the increasing Zr concentration. The higher the Zr content, the more complex the composition of inclusions in the alloy. The average diameter of inclusions in 0.004Zr steel was the smallest, which was 0.79 μm and the volume fraction was 0.018%. The highest yield strength, tensile strength, elongation, and impact energy of 0.004Zr alloy at room temperature were 548.3 MPa, 679.4 MPa, 25.7%, and 253.9 J. The structure of the TMPed steels was all tempered martensite. With the increase in tempering temperature, the yield and tensile strength of the experimental steel gradually decreased, while the elongation and impact energy gradually increased. The 0.004ZrD and 0.004ZrH alloys had the best yield strength and impact energy, which were 597.9 and 611.8 MPa and 225.9 and 243.3 J, respectively. In addition, the alloys showed good thermal stability during the aging at 600 °C for 1500 h. It was discovered that TMP is a simple and practical industrial technique that could successfully enhance the mechanical properties of CLAM steel without sacrificing impact toughness.
{"title":"Enhanced Mechanical Properties of CLAM by Zirconium Alloying and Thermo-Mechanical Processing","authors":"D. Zhan, Jihang Li, Dongwei Wang, Hui-shu Zhang, Guoxing Qiu, Yongkun Yang","doi":"10.3390/jne4010009","DOIUrl":"https://doi.org/10.3390/jne4010009","url":null,"abstract":"In this study, we present the effects of 0.004~0.098 wt% Zr and thermo-mechanical processing (TMP) on the microstructure and mechanical properties of the China RAFM steel, CLAM, as a feasibility study for improving mechanical properties. The inclusions in ingots were characterized using optical microscope (OM) and scanning electron microscope (SEM), which could be classified as fine simple particles and large complex particles. The complexity of the alloy’s inclusion composition increases with the increasing Zr concentration. The higher the Zr content, the more complex the composition of inclusions in the alloy. The average diameter of inclusions in 0.004Zr steel was the smallest, which was 0.79 μm and the volume fraction was 0.018%. The highest yield strength, tensile strength, elongation, and impact energy of 0.004Zr alloy at room temperature were 548.3 MPa, 679.4 MPa, 25.7%, and 253.9 J. The structure of the TMPed steels was all tempered martensite. With the increase in tempering temperature, the yield and tensile strength of the experimental steel gradually decreased, while the elongation and impact energy gradually increased. The 0.004ZrD and 0.004ZrH alloys had the best yield strength and impact energy, which were 597.9 and 611.8 MPa and 225.9 and 243.3 J, respectively. In addition, the alloys showed good thermal stability during the aging at 600 °C for 1500 h. It was discovered that TMP is a simple and practical industrial technique that could successfully enhance the mechanical properties of CLAM steel without sacrificing impact toughness.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"44 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-01-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89810476","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}
Maintenance of the DEMO breeding blanket includes the removal and replacement of plasma-facing components. To access the breeding blanket, multiple coolant pipes need to be removed to allow access to the tokamak. As an option to reduce downtime and increase maintenance speed, the pipe-connection concept is developed to allow the removal of multiple pipes at the same time using a remotely operated mechanical connection. The remotely operated multi-pipe Mechanical Pipe Connection (MPC) needs to fulfil multiple requirements, such as high operating temperature and high external forces while at the same time maintaining an acceptable level of sealing between the high-pressure fluid and vacuum surroundings. In addition to the external conditions, the pipes of multiple sizes and fluids are connected in a manifold configuration. Although this will reduce the overall time required to operate the mechanical pipe connection when compared to multiple single-pipe connections, this will introduce additional forces and stresses due the interaction between pipe flow (e.g., simultaneous high- and low-temperature fluid pipes on the same manifold) through the manifold flange. The requirements and the boundary conditions of the multi-pipe MPC are taken into consideration during the design process of MPC. The design process is carried out to find the optimum form and size to allow the mechanical function of the pipe connection during the maintenance phase while withstanding the extreme operating conditions that the MPC will face the during operational phase. The resulting design will then be analyzed using numerical methods to assess the capability of the MPC designs.
{"title":"Development of Mechanical Pipe-Connection Design for DEMO","authors":"Viktor Milushev, Azman Azka, M. Mittwollen","doi":"10.3390/jne4010008","DOIUrl":"https://doi.org/10.3390/jne4010008","url":null,"abstract":"Maintenance of the DEMO breeding blanket includes the removal and replacement of plasma-facing components. To access the breeding blanket, multiple coolant pipes need to be removed to allow access to the tokamak. As an option to reduce downtime and increase maintenance speed, the pipe-connection concept is developed to allow the removal of multiple pipes at the same time using a remotely operated mechanical connection. The remotely operated multi-pipe Mechanical Pipe Connection (MPC) needs to fulfil multiple requirements, such as high operating temperature and high external forces while at the same time maintaining an acceptable level of sealing between the high-pressure fluid and vacuum surroundings. In addition to the external conditions, the pipes of multiple sizes and fluids are connected in a manifold configuration. Although this will reduce the overall time required to operate the mechanical pipe connection when compared to multiple single-pipe connections, this will introduce additional forces and stresses due the interaction between pipe flow (e.g., simultaneous high- and low-temperature fluid pipes on the same manifold) through the manifold flange. The requirements and the boundary conditions of the multi-pipe MPC are taken into consideration during the design process of MPC. The design process is carried out to find the optimum form and size to allow the mechanical function of the pipe connection during the maintenance phase while withstanding the extreme operating conditions that the MPC will face the during operational phase. The resulting design will then be analyzed using numerical methods to assess the capability of the MPC designs.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-01-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75580757","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper presents the generation and the verification of the few-group homogenized cross-sections for Batan-Fuel. This code is routinely used for fuel management in the RSG-GAS. The Monte Carlo code Serpent 2 code, in conjunction with the latest nuclear data library ENDF/B-VIII.0, was used. Calculations using the existing newly generated few-group cross-section data were carried out for the 88th core. The calculated core parameters such as excess reactivity and control rod worth are compared to the available experimental data. On the other hand, the fuel burnup fraction and radial power peaking factor (PPF) are compared to the results of Serpent 2. It was shown that the new cross-section data have more consistency and a better agreement excess reactivity and control rod worth compared to the experimental data. Similarly, the U-235 burnup fraction and radial power peaking factor by the new cross-section data are also shown to concur well with Serpent 2. The newly generated few-group cross-sections are recommended to replace the existing ones for the fuel management of RSG-GAS.
{"title":"Improvement of Few-Group Homogenized Cross-Sections for RSG-GAS In-Core Fuel Management Code BATAN-FUEL","authors":"S. Pinem, D. Hartanto, Liem Peng Hong, W. Luthfi","doi":"10.1115/1.4056603","DOIUrl":"https://doi.org/10.1115/1.4056603","url":null,"abstract":"\u0000 This paper presents the generation and the verification of the few-group homogenized cross-sections for Batan-Fuel. This code is routinely used for fuel management in the RSG-GAS. The Monte Carlo code Serpent 2 code, in conjunction with the latest nuclear data library ENDF/B-VIII.0, was used. Calculations using the existing newly generated few-group cross-section data were carried out for the 88th core. The calculated core parameters such as excess reactivity and control rod worth are compared to the available experimental data. On the other hand, the fuel burnup fraction and radial power peaking factor (PPF) are compared to the results of Serpent 2. It was shown that the new cross-section data have more consistency and a better agreement excess reactivity and control rod worth compared to the experimental data. Similarly, the U-235 burnup fraction and radial power peaking factor by the new cross-section data are also shown to concur well with Serpent 2. The newly generated few-group cross-sections are recommended to replace the existing ones for the fuel management of RSG-GAS.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"225 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-01-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86008692","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}
I. Jõgi, P. Paris, E. Bernard, M. Diez, E. Tsitrone, A. Hakola, J. Likonen, T. Vuoriheimo, E. Grigore
Fuel retention monitoring in tokamak walls requires the development of remote composition analysis methods such as laser-induced breakdown spectroscopy (LIBS). The present study investigates the feasibility of the LIBS method to analyse the composition and fuel retention in three samples from WEST divertor erosion marker tiles after the experimental campaign C3. The investigated samples originated from tile regions outside of strong erosion and deposition regions, where the variation of thin deposit layers is relatively small and facilitates cross-comparison between different analysis methods. The depth profiles of main constituents W, Mo and C were consistent with depth profiles determined by other composition analysis methods, such as glow-discharge optical emission spectroscopy (GDOES) and secondary ion mass spectrometry (SIMS). The average LIBS depth resolution determined from depth profiles was 100 nm/shot. The averaging of the spectra collected from multiple spots of a same sample allowed us to improve the signal-to-noise ratio, investigate the presence of fuel D and trace impurities such as O and B. In the investigated tile regions with negligible erosion and deposition, these impurities were clearly detectable during the first laser shot, while the signal decreased to noise level after a few subsequent laser shots at the same spot. LIBS investigation of samples originating from the deposition regions of tiles may further clarify LIBS’ ability to investigate trace impurities.
{"title":"Ex Situ LIBS Analysis of WEST Divertor Wall Tiles after C3 Campaign","authors":"I. Jõgi, P. Paris, E. Bernard, M. Diez, E. Tsitrone, A. Hakola, J. Likonen, T. Vuoriheimo, E. Grigore","doi":"10.3390/jne4010007","DOIUrl":"https://doi.org/10.3390/jne4010007","url":null,"abstract":"Fuel retention monitoring in tokamak walls requires the development of remote composition analysis methods such as laser-induced breakdown spectroscopy (LIBS). The present study investigates the feasibility of the LIBS method to analyse the composition and fuel retention in three samples from WEST divertor erosion marker tiles after the experimental campaign C3. The investigated samples originated from tile regions outside of strong erosion and deposition regions, where the variation of thin deposit layers is relatively small and facilitates cross-comparison between different analysis methods. The depth profiles of main constituents W, Mo and C were consistent with depth profiles determined by other composition analysis methods, such as glow-discharge optical emission spectroscopy (GDOES) and secondary ion mass spectrometry (SIMS). The average LIBS depth resolution determined from depth profiles was 100 nm/shot. The averaging of the spectra collected from multiple spots of a same sample allowed us to improve the signal-to-noise ratio, investigate the presence of fuel D and trace impurities such as O and B. In the investigated tile regions with negligible erosion and deposition, these impurities were clearly detectable during the first laser shot, while the signal decreased to noise level after a few subsequent laser shots at the same spot. LIBS investigation of samples originating from the deposition regions of tiles may further clarify LIBS’ ability to investigate trace impurities.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"31 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-01-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78039780","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}
Neural networks require a large quantity of training spectra and detector responses in order to learn to solve the inverse problem of neutron spectrum unfolding. In addition, due to the under-determined nature of unfolding, non-physical spectra which would not be encountered in usage should not be included in the training set. While physically realistic training spectra are commonly determined experimentally or generated through Monte Carlo simulation, this can become prohibitively expensive when considering the quantity of spectra needed to effectively train an unfolding network. In this paper, we present three algorithms for the generation of large quantities of realistic and physically motivated neutron energy spectra. Using an IAEA compendium of 251 spectra, we compare the unfolding performance of neural networks trained on spectra from these algorithms, when unfolding real-world spectra, to two baselines. We also investigate general methods for evaluating the performance of and optimizing feature engineering algorithms.
{"title":"Data Augmentation for Neutron Spectrum Unfolding with Neural Networks","authors":"James McGreivy, J. Manfredi, Daniel Siefman","doi":"10.3390/jne4010006","DOIUrl":"https://doi.org/10.3390/jne4010006","url":null,"abstract":"Neural networks require a large quantity of training spectra and detector responses in order to learn to solve the inverse problem of neutron spectrum unfolding. In addition, due to the under-determined nature of unfolding, non-physical spectra which would not be encountered in usage should not be included in the training set. While physically realistic training spectra are commonly determined experimentally or generated through Monte Carlo simulation, this can become prohibitively expensive when considering the quantity of spectra needed to effectively train an unfolding network. In this paper, we present three algorithms for the generation of large quantities of realistic and physically motivated neutron energy spectra. Using an IAEA compendium of 251 spectra, we compare the unfolding performance of neural networks trained on spectra from these algorithms, when unfolding real-world spectra, to two baselines. We also investigate general methods for evaluating the performance of and optimizing feature engineering algorithms.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-01-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75220362","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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