M. Brovchenko, J. Kloosterman, L. Luzzi, E. Merle, D. Heuer, A. Laureau, O. Feynberg, V. Ignatiev, M. Aufiero, A. Cammi, C. Fiorina, F. Alcaro, S. Dulla, P. Ravetto, Lodewijk Frima, D. Lathouwers, B. Merk
This paper describes the neutronic benchmarks and the results obtained by the various participants of the FP7 project EVOL and the ROSATOM project MARS. The aim of the benchmarks was two-fold: first to verify and validate each of the code packages of the project partners, adapted for liquid-fueled reactors, and second to check the dependence of the core characteristics to nuclear data set for application on a molten salt fast reactor (MSFR). The MSFR operates with the thorium fuel cycle and can be started with 233U-enriched U and/or TRU elements as initial fissile load. All three compositions were covered by the present benchmark. The calculations have confirmed that the MSFR has very favorable characteristics not present in other Gen4 fast reactors, like strong negative temperature and void reactivity coefficients, a low-fissile inventory, a reduced long-lived waste production and its burning capacities of nuclear waste produced in currently operational reactors.
{"title":"Neutronic benchmark of the molten salt fast reactor in the frame of the EVOL and MARS collaborative projects","authors":"M. Brovchenko, J. Kloosterman, L. Luzzi, E. Merle, D. Heuer, A. Laureau, O. Feynberg, V. Ignatiev, M. Aufiero, A. Cammi, C. Fiorina, F. Alcaro, S. Dulla, P. Ravetto, Lodewijk Frima, D. Lathouwers, B. Merk","doi":"10.1051/EPJN/2018052","DOIUrl":"https://doi.org/10.1051/EPJN/2018052","url":null,"abstract":"This paper describes the neutronic benchmarks and the results obtained by the various participants of the FP7 project EVOL and the ROSATOM project MARS. The aim of the benchmarks was two-fold: first to verify and validate each of the code packages of the project partners, adapted for liquid-fueled reactors, and second to check the dependence of the core characteristics to nuclear data set for application on a molten salt fast reactor (MSFR). The MSFR operates with the thorium fuel cycle and can be started with 233U-enriched U and/or TRU elements as initial fissile load. All three compositions were covered by the present benchmark. The calculations have confirmed that the MSFR has very favorable characteristics not present in other Gen4 fast reactors, like strong negative temperature and void reactivity coefficients, a low-fissile inventory, a reduced long-lived waste production and its burning capacities of nuclear waste produced in currently operational reactors.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":" ","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/EPJN/2018052","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47130447","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A-H Puichaud, C. Flament, Aziz Chniouel, F. Lomello, E. Rouesne, P. Giroux, H. Maskrot, F. Schuster, J. Béchade
Additive manufacturing (AM) is rapidly expanding in many industrial applications because of the versatile possibilities of fast and complex fabrication of added value products. This manufacturing process would significantly reduce manufacturing time and development cost for nuclear components. However, the process leads to materials with complex microstructures, and their structural stability for nuclear application is still uncertain. This study focuses on 316L stainless steel fabricated by selective laser melting (SLM) in the context of nuclear application, and compares with a cold-rolled solution annealed 316L sample. The effect of heat treatment (HT) and hot isostatic pressing (HIP) on the microstructure and mechanical properties is discussed. It was found that after HT, the material microstructure remains mostly unchanged, while the HIP treatment removes the materials porosity, and partially re-crystallises the microstructure. Finally, the tensile tests showed excellent results, satisfying RCC-MR code requirements for all AM materials.
{"title":"Microstructure and mechanical properties relationship of additively manufactured 316L stainless steel by selective laser melting","authors":"A-H Puichaud, C. Flament, Aziz Chniouel, F. Lomello, E. Rouesne, P. Giroux, H. Maskrot, F. Schuster, J. Béchade","doi":"10.1051/epjn/2019051","DOIUrl":"https://doi.org/10.1051/epjn/2019051","url":null,"abstract":"Additive manufacturing (AM) is rapidly expanding in many industrial applications because of the versatile possibilities of fast and complex fabrication of added value products. This manufacturing process would significantly reduce manufacturing time and development cost for nuclear components. However, the process leads to materials with complex microstructures, and their structural stability for nuclear application is still uncertain. This study focuses on 316L stainless steel fabricated by selective laser melting (SLM) in the context of nuclear application, and compares with a cold-rolled solution annealed 316L sample. The effect of heat treatment (HT) and hot isostatic pressing (HIP) on the microstructure and mechanical properties is discussed. It was found that after HT, the material microstructure remains mostly unchanged, while the HIP treatment removes the materials porosity, and partially re-crystallises the microstructure. Finally, the tensile tests showed excellent results, satisfying RCC-MR code requirements for all AM materials.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":" ","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/epjn/2019051","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49304950","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 reactor accidents that involve rapid overheating of oxide fuel, overpressurization of gas-filled bubbles and pores may lead to rupture of these cavities, fine fragmentation of the fuel material, and burst-type release of the cavity gas. Analytical rupture criteria for various types of cavities exist, but application of these criteria requires that microstructural characteristics of the fuel, such as cavity size, shape and number density, are known together with the gas content of the cavities. In this paper, we integrate rupture criteria for two kinds of cavities with models that calculate the aforementioned parameters in UO2 LWR fuel for a given operating history. The models are intended for implementation in engineering type computer programs for thermal-mechanical analyses of LWR fuel rods. Here, they have been implemented in the FRAPCON and FRAPTRAN programs and validated against experiments that simulate LOCA and RIA conditions. The capabilities and shortcomings of the proposed models are discussed in light of selected results from this validation. Calculated results suggest that the extent of fuel fragmentation and transient fission gas release depends strongly on the pre-accident fuel microstructure and fission gas distribution, but also on rapid changes in the external pressure exerted on the fuel pellets during the accident.
{"title":"Modelling of fine fragmentation and fission gas release of UO2 fuel in accident conditions","authors":"L. Jernkvist","doi":"10.1051/epjn/2019030","DOIUrl":"https://doi.org/10.1051/epjn/2019030","url":null,"abstract":"In reactor accidents that involve rapid overheating of oxide fuel, overpressurization of gas-filled bubbles and pores may lead to rupture of these cavities, fine fragmentation of the fuel material, and burst-type release of the cavity gas. Analytical rupture criteria for various types of cavities exist, but application of these criteria requires that microstructural characteristics of the fuel, such as cavity size, shape and number density, are known together with the gas content of the cavities. In this paper, we integrate rupture criteria for two kinds of cavities with models that calculate the aforementioned parameters in UO2 LWR fuel for a given operating history. The models are intended for implementation in engineering type computer programs for thermal-mechanical analyses of LWR fuel rods. Here, they have been implemented in the FRAPCON and FRAPTRAN programs and validated against experiments that simulate LOCA and RIA conditions. The capabilities and shortcomings of the proposed models are discussed in light of selected results from this validation. Calculated results suggest that the extent of fuel fragmentation and transient fission gas release depends strongly on the pre-accident fuel microstructure and fission gas distribution, but also on rapid changes in the external pressure exerted on the fuel pellets during the accident.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":" ","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/epjn/2019030","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45851412","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 study, conceptual design for primary heat exchanger of the Molten Salt Fast Reactor is made. The design was carried out to remove the produced heat from the reactor developed under the SAMOFAR project. Nominal power of the reactor is 3 GWth and it has 16 heat exchangers. There are several requirements related to the heat exchanger. To sustain the steady-state conditions, heat exchangers have to transfer the heat produced in the core and it has to maintain the temperature drop as much as the temperature rise in the core due to the fission. It should do it as fast as possible. It must also ensure that the fuel temperature does not reach the freezing temperature to avoid solidification. In doing so, the fuel volume in the heat exchanger must not exceed the specified limit. Design studies were carried out taking into account all requirements and final geometric configurations were determined. Plate type heat exchanger was adopted in this study. 3D CFD analyses were performed to investigate the thermal-hydraulic behavior of the system. Analyses were made by ANSYS-Fluent commercial code. Results are in a good agreement with limitations and requirements specified for the reactor designed under the SAMOFAR project.
{"title":"Heat exchanger design studies for molten salt fast reactor","authors":"Uğur Köse, U. Koç, L. Erbay, E. Öğüt, H. Ayhan","doi":"10.1051/epjn/2019032","DOIUrl":"https://doi.org/10.1051/epjn/2019032","url":null,"abstract":"In this study, conceptual design for primary heat exchanger of the Molten Salt Fast Reactor is made. The design was carried out to remove the produced heat from the reactor developed under the SAMOFAR project. Nominal power of the reactor is 3 GWth and it has 16 heat exchangers. There are several requirements related to the heat exchanger. To sustain the steady-state conditions, heat exchangers have to transfer the heat produced in the core and it has to maintain the temperature drop as much as the temperature rise in the core due to the fission. It should do it as fast as possible. It must also ensure that the fuel temperature does not reach the freezing temperature to avoid solidification. In doing so, the fuel volume in the heat exchanger must not exceed the specified limit. Design studies were carried out taking into account all requirements and final geometric configurations were determined. Plate type heat exchanger was adopted in this study. 3D CFD analyses were performed to investigate the thermal-hydraulic behavior of the system. Analyses were made by ANSYS-Fluent commercial code. Results are in a good agreement with limitations and requirements specified for the reactor designed under the SAMOFAR project.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":" ","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/epjn/2019032","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42553335","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 a liquid fuel molten salt reactor (MSR) a key factor to consider upon its design is the strong coupling between different physics present such as neutronics, thermo-mechanics and thermal-hydraulics. Focusing in the thermal-hydraulics aspect, it is required that the heat transfer is well characterized. For this purpose, turbulent models used for FLiNaK flow must be valid, and its thermophysical properties must be accurately described. In the literature, there are several expressions for each material property, with differences that can be significant. The goal of this study is to demonstrate and quantify the impact that the uncertainty in thermophysical properties has on key metrics of thermal hydraulic importance for MSRs, in particular on the heat transfer coefficient. In order to achieve this, computational fluid dynamics (CFD) simulations using the RANS k-ω SST model were compared to published experiment data on molten salt. Various correlations for FLiNaK’s material properties were used. It was observed that the uncertainty in FLiNaK’s thermophysical properties lead to a significant variance in the heat coefficient. Motivated by this, additional CFD simulations were done to obtain sensitivity coefficients for each thermophysical property. With this information, the effect of the variation of each one of the material properties on the heat transfer coefficient was quantified performing a one factor at a time approach (OAT). The results of this sensitivity analysis showed that the most critical thermophysical properties of FLiNaK towards the determination of the heat transfer coefficient are the viscosity and the thermal conductivity. More specifically the dimensionless sensitivity coefficient, which is defined as the percent variation of the heat transfer with respect to the percent variation of the respective property, was −0.51 and 0.67 respectively. According to the different correlations, the maximum percent variations for these properties is 18% and 26% respectively, which yields a variation in the predicted heat transfer coefficient as high as 9% and 17% for the viscosity and thermal conductivity, respectively. It was also demonstrated that the Nusselt number trends found from the simulations were captured much better using the Sieder Tate correlation than the Dittus Boelter correlation. Future work accommodating additional turbulence models and higher fidelity physics will help to determine whether the Sieder Tate expression truly captures the physics of interest or whether the agreement seen in the current work is simply reflective of the single turbulence model employed.
{"title":"Influence of molten salt-(FLiNaK) thermophysical properties on a heated tube using CFD RANS turbulence modeling of an experimental testbed","authors":"Ramiro Freile, M. Kimber","doi":"10.1051/epjn/2019027","DOIUrl":"https://doi.org/10.1051/epjn/2019027","url":null,"abstract":"In a liquid fuel molten salt reactor (MSR) a key factor to consider upon its design is the strong coupling between different physics present such as neutronics, thermo-mechanics and thermal-hydraulics. Focusing in the thermal-hydraulics aspect, it is required that the heat transfer is well characterized. For this purpose, turbulent models used for FLiNaK flow must be valid, and its thermophysical properties must be accurately described. In the literature, there are several expressions for each material property, with differences that can be significant. The goal of this study is to demonstrate and quantify the impact that the uncertainty in thermophysical properties has on key metrics of thermal hydraulic importance for MSRs, in particular on the heat transfer coefficient. In order to achieve this, computational fluid dynamics (CFD) simulations using the RANS k-ω SST model were compared to published experiment data on molten salt. Various correlations for FLiNaK’s material properties were used. It was observed that the uncertainty in FLiNaK’s thermophysical properties lead to a significant variance in the heat coefficient. Motivated by this, additional CFD simulations were done to obtain sensitivity coefficients for each thermophysical property. With this information, the effect of the variation of each one of the material properties on the heat transfer coefficient was quantified performing a one factor at a time approach (OAT). The results of this sensitivity analysis showed that the most critical thermophysical properties of FLiNaK towards the determination of the heat transfer coefficient are the viscosity and the thermal conductivity. More specifically the dimensionless sensitivity coefficient, which is defined as the percent variation of the heat transfer with respect to the percent variation of the respective property, was −0.51 and 0.67 respectively. According to the different correlations, the maximum percent variations for these properties is 18% and 26% respectively, which yields a variation in the predicted heat transfer coefficient as high as 9% and 17% for the viscosity and thermal conductivity, respectively. It was also demonstrated that the Nusselt number trends found from the simulations were captured much better using the Sieder Tate correlation than the Dittus Boelter correlation. Future work accommodating additional turbulence models and higher fidelity physics will help to determine whether the Sieder Tate expression truly captures the physics of interest or whether the agreement seen in the current work is simply reflective of the single turbulence model employed.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/epjn/2019027","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41889276","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}
F. Rouillard, B. Duprey, J. Courouau, R. Robin, P. Aubry, C. Blanc, M. Tabarant, H. Maskrot, L. Nicolas, M. Blat-Yrieix, G. Rolland, T. Marlaud
The feedback produced by operating Sodium-cooled Fast Reactors (SFRs) has shown the importance of material tribological properties. Where galling or adhesive wear cannot be allowed, hardfacing alloys, known to be galling-resistant coatings, are usually applied on rubbing surfaces. The most used coating is the cobalt-base alloy named Stellite 6® because of its outstanding friction and wear behaviour. Nevertheless, cobalt is an element which activates in the reactor leading to complex management of safety during reactor maintenance and mainly decommissioning. As a consequence, a collaborative work between CEA, EDF and FRAMATOME has been launched for selecting promising cobalt-free hardfacing alloys for the 600 MWe Sodium-cooled Fast breeder reactor project named ASTRID. Several nickel-base alloys have been selected from literature review then deposited by Plasma Transferred Arc or Laser Cladding on 17Cr austenitic stainless steel 316L(N) according to RCC-MRx Code (AFCEN Code). Among the numerous properties required for qualifying their use as hardfacing alloys in SFR, good corrosion behaviour and good friction and wear behaviour in sodium are essential. First results on these properties are shown in this article. Firstly, the corrosion behaviour of all coatings was evaluated through exposure tests in purified sodium for 5000 h at 400 °C. All coatings showed an acceptable corrosion behaviour in sodium. Finally, the friction and wear properties of one alloy candidate, NiCrBSi alloy, were studied in sodium in a dedicated designed facility. The influence of the oxygen concentration in sodium on the friction and wear properties was evaluated.
{"title":"Evaluation of cobalt free coatings as hardfacing material candidates in sodium-cooled fast reactor and effect of oxygen in sodium on the tribological behaviour","authors":"F. Rouillard, B. Duprey, J. Courouau, R. Robin, P. Aubry, C. Blanc, M. Tabarant, H. Maskrot, L. Nicolas, M. Blat-Yrieix, G. Rolland, T. Marlaud","doi":"10.1051/epjn/2019025","DOIUrl":"https://doi.org/10.1051/epjn/2019025","url":null,"abstract":"The feedback produced by operating Sodium-cooled Fast Reactors (SFRs) has shown the importance of material tribological properties. Where galling or adhesive wear cannot be allowed, hardfacing alloys, known to be galling-resistant coatings, are usually applied on rubbing surfaces. The most used coating is the cobalt-base alloy named Stellite 6® because of its outstanding friction and wear behaviour. Nevertheless, cobalt is an element which activates in the reactor leading to complex management of safety during reactor maintenance and mainly decommissioning. As a consequence, a collaborative work between CEA, EDF and FRAMATOME has been launched for selecting promising cobalt-free hardfacing alloys for the 600 MWe Sodium-cooled Fast breeder reactor project named ASTRID. Several nickel-base alloys have been selected from literature review then deposited by Plasma Transferred Arc or Laser Cladding on 17Cr austenitic stainless steel 316L(N) according to RCC-MRx Code (AFCEN Code). Among the numerous properties required for qualifying their use as hardfacing alloys in SFR, good corrosion behaviour and good friction and wear behaviour in sodium are essential. First results on these properties are shown in this article. Firstly, the corrosion behaviour of all coatings was evaluated through exposure tests in purified sodium for 5000 h at 400 °C. All coatings showed an acceptable corrosion behaviour in sodium. Finally, the friction and wear properties of one alloy candidate, NiCrBSi alloy, were studied in sodium in a dedicated designed facility. The influence of the oxygen concentration in sodium on the friction and wear properties was evaluated.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/epjn/2019025","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826174","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}
With Europe's ageing fleet of nuclear reactors operating closer to their safety limits, the monitoring of such reactors through complex models has become of great interest to maintain a high level of availability and safety. Therefore, we propose an extended Deep Learning framework as part of the CORTEX Horizon 2020 EU project for the unfolding of reactor transfer functions from induced neutron noise sources. The unfolding allows for the identification and localisation of reactor core perturbation sources from neutron detector readings in Pressurised Water Reactors. A 3D Convolutional Neural Network (3D-CNN) and Long Short-Term Memory (LSTM) Recurrent Neural Network (RNN) have been presented, each to study the signals presented in frequency and time domain respectively. The proposed approach achieves state-of-the-art results with the classification of perturbation type in the frequency domain reaching 99.89% accuracy and localisation of the classified perturbation source being regressed to 0.2902 Mean Absolute Error (MAE).
{"title":"3D convolutional and recurrent neural networks for reactor perturbation unfolding and anomaly detection","authors":"A. Durrant, G. Leontidis, S. Kollias","doi":"10.1051/EPJN/2019047","DOIUrl":"https://doi.org/10.1051/EPJN/2019047","url":null,"abstract":"With Europe's ageing fleet of nuclear reactors operating closer to their safety limits, the monitoring of such reactors through complex models has become of great interest to maintain a high level of availability and safety. Therefore, we propose an extended Deep Learning framework as part of the CORTEX Horizon 2020 EU project for the unfolding of reactor transfer functions from induced neutron noise sources. The unfolding allows for the identification and localisation of reactor core perturbation sources from neutron detector readings in Pressurised Water Reactors. A 3D Convolutional Neural Network (3D-CNN) and Long Short-Term Memory (LSTM) Recurrent Neural Network (RNN) have been presented, each to study the signals presented in frequency and time domain respectively. The proposed approach achieves state-of-the-art results with the classification of perturbation type in the frequency domain reaching 99.89% accuracy and localisation of the classified perturbation source being regressed to 0.2902 Mean Absolute Error (MAE).","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":" ","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-06-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/EPJN/2019047","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47341775","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}
Obtaining nuclear data is an international activity with new and updated data constantly being determined by thousands of scientists at major research centres worldwide. Because of the large amounts of data generated and the formats used to store these data, the field of nuclear data is highly specialised. To make the most important key data more accessible to a wider audience, nuclide charts have been developed. In this article, we present the scientific highlights of the new 10th Edition of the Karlsruhe Nuclide Chart. The main focus of this Chart is to provide structured, accurate information on the half-lives and decay modes, as well as energies of the emitted radiation for over 4000 experimentally observed ground states and isomer nuclides to an interdisciplinary audience.
{"title":"Karlsruhe Nuclide Chart – New 10th edition 2018 ","authors":"Z. Sóti, J. Magill, R. Dreher","doi":"10.1051/EPJN/2019004","DOIUrl":"https://doi.org/10.1051/EPJN/2019004","url":null,"abstract":"Obtaining nuclear data is an international activity with new and updated data constantly being determined by thousands of scientists at major research centres worldwide. Because of the large amounts of data generated and the formats used to store these data, the field of nuclear data is highly specialised. To make the most important key data more accessible to a wider audience, nuclide charts have been developed. In this article, we present the scientific highlights of the new 10th Edition of the Karlsruhe Nuclide Chart. The main focus of this Chart is to provide structured, accurate information on the half-lives and decay modes, as well as energies of the emitted radiation for over 4000 experimentally observed ground states and isomer nuclides to an interdisciplinary audience.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":" ","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/EPJN/2019004","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42597852","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}
Coarse mesh nodal methods are widely used in the analysis of nuclear reactors. However, these methods provide only average values of the neutron fluxes. From a safety point of view, it is important to have an accurate analysis of the pin to pin flux distribution that nodal methods are not able to provide. Many articles have been published that make use of mathematical techniques to determine flux distributions. Most of these techniques use expansion functions to estimate these distributions. The expansion coefficients of these works are determined by conditions that take into account the average values of certain fluxes supplied by the nodal methods. There are also methods that employ analytical solutions of the neutron diffusion equation. This article presents a different approach for calculating the pin to pin neutron flux distribution for a PWR reactor. The developed method uses support vector regression (SVR) technique to determine this pin to pin neutron flux. The SVR technique uses average data computed with the Nodal Expansion Method (NEM) for learning purposes. A total of 70% of the computed data were used for training and 30% for validation, using multifold-cross-validation. Two fuel elements were removed from the training and validation sets, to test the method. Less than 2% errors were found when compared to the values obtained by the nodal expansion method (NEM), using a fine-mesh spatial discretization. We concluded that use of SVR to reconstruct pin to pin fluxes is another option, which will be of great value in fuel reload calculations, since the same parameters will be applied to all cycles, thus expediting calculations when compared to standard procedure calculations.
{"title":"Pin to pin neutron flux reconstruction in a PWR reactor using support vector regression (SVR) technique","authors":"W.F.P. Neto, A. Alvim, F. Silva, L. Alvim","doi":"10.1051/EPJN/2018051","DOIUrl":"https://doi.org/10.1051/EPJN/2018051","url":null,"abstract":"Coarse mesh nodal methods are widely used in the analysis of nuclear reactors. However, these methods provide only average values of the neutron fluxes. From a safety point of view, it is important to have an accurate analysis of the pin to pin flux distribution that nodal methods are not able to provide. Many articles have been published that make use of mathematical techniques to determine flux distributions. Most of these techniques use expansion functions to estimate these distributions. The expansion coefficients of these works are determined by conditions that take into account the average values of certain fluxes supplied by the nodal methods. There are also methods that employ analytical solutions of the neutron diffusion equation. This article presents a different approach for calculating the pin to pin neutron flux distribution for a PWR reactor. The developed method uses support vector regression (SVR) technique to determine this pin to pin neutron flux. The SVR technique uses average data computed with the Nodal Expansion Method (NEM) for learning purposes. A total of 70% of the computed data were used for training and 30% for validation, using multifold-cross-validation. Two fuel elements were removed from the training and validation sets, to test the method. Less than 2% errors were found when compared to the values obtained by the nodal expansion method (NEM), using a fine-mesh spatial discretization. We concluded that use of SVR to reconstruct pin to pin fluxes is another option, which will be of great value in fuel reload calculations, since the same parameters will be applied to all cycles, thus expediting calculations when compared to standard procedure calculations.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/EPJN/2018051","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57825897","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 examines how intermediaries could interact with other important actors identified by the multi-level perspective (MLP) framework, the niche actors and regime actors, to create niches for nuclear heat production in France. Whatever is the source, recovering the wasted heat is a matter of energy efficiency. Nuclear plants could remain used for several decades in France. It is thus legitimate to investigate those possible niche nurturing processes which may allow a more efficient use of this technology. Challenges are high, and our conclusions modest regarding the possible breaking through of such exploratory and collective systems. Without significant windows of opportunity, even the most willing intermediation may not be able to change the status quo. It is however important to highlight the multifarious pathways that energy transitions could follow. Drawing on lessons from the MLP, this paper proposes three key actions for intermediation willing to move beyond technology-push approaches that can lead to tension and low legitimacy. These are, sharing questions instead of knowledge; mobilise, interest, involve a legitimate place; and prevent or avoid conflicts among stakeholders. Regime changes possibly enhancing the deployment of sustainable heating systems, not only nuclear plant sourced, are also discussed.
{"title":"Stimulating niche nurturing process for heat production with nuclear plants in France: A multi-level perspective","authors":"Martin Leurent","doi":"10.1051/EPJN/2019001","DOIUrl":"https://doi.org/10.1051/EPJN/2019001","url":null,"abstract":"This paper examines how intermediaries could interact with other important actors identified by the multi-level perspective (MLP) framework, the niche actors and regime actors, to create niches for nuclear heat production in France. Whatever is the source, recovering the wasted heat is a matter of energy efficiency. Nuclear plants could remain used for several decades in France. It is thus legitimate to investigate those possible niche nurturing processes which may allow a more efficient use of this technology. Challenges are high, and our conclusions modest regarding the possible breaking through of such exploratory and collective systems. Without significant windows of opportunity, even the most willing intermediation may not be able to change the status quo. It is however important to highlight the multifarious pathways that energy transitions could follow. Drawing on lessons from the MLP, this paper proposes three key actions for intermediation willing to move beyond technology-push approaches that can lead to tension and low legitimacy. These are, sharing questions instead of knowledge; mobilise, interest, involve a legitimate place; and prevent or avoid conflicts among stakeholders. Regime changes possibly enhancing the deployment of sustainable heating systems, not only nuclear plant sourced, are also discussed.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1051/EPJN/2019001","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57825907","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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