Dimitar G. Chereshkov, Mikhail Yu. Ternovykh, Georgiy V. Tikhomirov, Alexander A. Ryzhkov
The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO 2 , MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead- and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.
{"title":"Nuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison","authors":"Dimitar G. Chereshkov, Mikhail Yu. Ternovykh, Georgiy V. Tikhomirov, Alexander A. Ryzhkov","doi":"10.3897/nucet.9.111919","DOIUrl":"https://doi.org/10.3897/nucet.9.111919","url":null,"abstract":"The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO 2 , MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead- and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"29 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135617701","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}
Yury A. Kazansky, Nikita O. Kushnir, Ekaterina S. Khnykina
This paper considers the use of unconventional fuel in nuclear power reactors, using the example of a VVER-type unit, in order to find out the possibility of saving natural fissile uranium nuclei. Saving fissile uranium is one of the important tasks, the solution of which will give time for the development of a two-component nuclear power industry that will have no problems with fuel resources. However, at present, the reserves of cheap uranium can provide the existing level of global nuclear energy for only 80–100 years. The main components of this proposed fuel are 232 Th and fissile isotopes of uranium: 235 U (loaded) and 233 U (produced from thorium). All the uranium isotopes and added 235 U nuclei at the beginning of the campaign account for about 6% of the number of thorium nuclei and uranium isotopes. The abbreviated name of this fuel is TORUR-5. To keep fissionable nuclei in the fuel cycle after the spent fuel is unloaded, it is envisaged that all the heavy nuclei will be returned back to the reactor after they have been cleaned from fission fragments, i.e., the fuel cycle will be closed. At the same time, the principle of annual movement of fuel assemblies (as they burn up) is the same as in the existing VVER-1000 reactors. Using the Serpent software, a reactor model was built, the composition and dimensions of which were close to the parameters of the VVER-1000 serial unit. The main results of calculations were the quantitative compositions of isotopes annually loaded into the reactor as well as the amounts of 235 U and thorium added also annually. The analysis of the obtained results allowed us to make the following conclusions. The annual reloading of 235 U during the computation period is required almost at a constant level and, in comparison with uranium fuel, is about half as much. This is feasible for the following reasons. Part of the fissions of 235 U is replaced by the fission of 233 U produced from 232 Th. In addition, fissionable nuclei are kept in the closed Th-U fuel cycle. This is the first “advantage” of the proposed fuel. TORUR-5 requires uranium enriched to at least 90%, the cost of which is several times higher than that of 3–5% enriched uranium. But since much less highly enriched uranium is required, the cost of fuel for a TORUR-5-fueled VVER-1000 reactor is significantly lower. This is the second “advantage” of the proposed fuel. The negative characteristic of TORUR-5, which requires further investigation, is that, after the initial loading, several uranium isotopes appear in the returned fuel, the total radioactivity of which, according to estimates, exceeds the radioactivity of traditional 3–5% enriched uranium fuel by several thousand times. At the same time, the radioactivity of discharged spent conventional fuel exceeds the radioactivity of fresh fuel by millions of times, and this problem has been solved at NPPs both organizationally and technically. Therefore, it will be necessary to develop a technology for
{"title":"Multiple usage of thorium-based fuel in a VVER-1000 reactor","authors":"Yury A. Kazansky, Nikita O. Kushnir, Ekaterina S. Khnykina","doi":"10.3897/nucet.9.101762","DOIUrl":"https://doi.org/10.3897/nucet.9.101762","url":null,"abstract":"This paper considers the use of unconventional fuel in nuclear power reactors, using the example of a VVER-type unit, in order to find out the possibility of saving natural fissile uranium nuclei. Saving fissile uranium is one of the important tasks, the solution of which will give time for the development of a two-component nuclear power industry that will have no problems with fuel resources. However, at present, the reserves of cheap uranium can provide the existing level of global nuclear energy for only 80–100 years. The main components of this proposed fuel are 232 Th and fissile isotopes of uranium: 235 U (loaded) and 233 U (produced from thorium). All the uranium isotopes and added 235 U nuclei at the beginning of the campaign account for about 6% of the number of thorium nuclei and uranium isotopes. The abbreviated name of this fuel is TORUR-5. To keep fissionable nuclei in the fuel cycle after the spent fuel is unloaded, it is envisaged that all the heavy nuclei will be returned back to the reactor after they have been cleaned from fission fragments, i.e., the fuel cycle will be closed. At the same time, the principle of annual movement of fuel assemblies (as they burn up) is the same as in the existing VVER-1000 reactors. Using the Serpent software, a reactor model was built, the composition and dimensions of which were close to the parameters of the VVER-1000 serial unit. The main results of calculations were the quantitative compositions of isotopes annually loaded into the reactor as well as the amounts of 235 U and thorium added also annually. The analysis of the obtained results allowed us to make the following conclusions. The annual reloading of 235 U during the computation period is required almost at a constant level and, in comparison with uranium fuel, is about half as much. This is feasible for the following reasons. Part of the fissions of 235 U is replaced by the fission of 233 U produced from 232 Th. In addition, fissionable nuclei are kept in the closed Th-U fuel cycle. This is the first “advantage” of the proposed fuel. TORUR-5 requires uranium enriched to at least 90%, the cost of which is several times higher than that of 3–5% enriched uranium. But since much less highly enriched uranium is required, the cost of fuel for a TORUR-5-fueled VVER-1000 reactor is significantly lower. This is the second “advantage” of the proposed fuel. The negative characteristic of TORUR-5, which requires further investigation, is that, after the initial loading, several uranium isotopes appear in the returned fuel, the total radioactivity of which, according to estimates, exceeds the radioactivity of traditional 3–5% enriched uranium fuel by several thousand times. At the same time, the radioactivity of discharged spent conventional fuel exceeds the radioactivity of fresh fuel by millions of times, and this problem has been solved at NPPs both organizationally and technically. Therefore, it will be necessary to develop a technology for ","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"28 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135187959","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 paper presents the results of modeling of changes in the isotopic composition of the liquid-salt fuel circulating in the experimental channel of the MBIR reactor facility. The authors tested the ISTAR software environment adapted for solving burnup equations in problems with variable power levels. The loop channel parameters, including two heat exchanger options, were estimated to obtain the appropriate salt transit time through the loop channel zones. Two problems of a circulating fuel system (loop) modeling are considered, namely: (1) modeling the equilibrium salt isotope composition in such a system; and (2) developing a technique for modeling nonstationary isotope kinetics in the MBIR reactor loop. Non-stationary isotope kinetics can be modeled as sequential burnup of nuclides in the neutron field and decay during movement in the external circuit. The authors also developed an algorithm for modeling changes in the isotopic composition of fuel salt during its circulation, taking into account the sequential transfer of a given salt volume from the burnup zone to the zone outside the reactor core. Based on this algorithm, a software package was created using the Python 3.9 programming language and ISTAR modules. In addition, a description of the calculation methodology was given and some calculation results obtained using the software were presented. In the process of working with the program, it was found that, for the given times of the fuel being in each of the zones (2 and 200 seconds, respectively), modeling the change in the isotopic composition during the fuel campaign (500 days) will require the calculation of more than 500 thousand steps. In order to save time, it is necessary to find out whether it will be possible to reduce the number of calls to the neutronic calculation code due to a slight change in the isotopic composition of the fuel in the loop per one burnup step. Work is currently underway to optimize this process.
{"title":"Isotope kinetics modeling in a circulating fuel system: a case study of the MBIR reactor loop","authors":"D. S. Kuzenkova, Victor Yu. Blandinskiy","doi":"10.3897/nucet.9.107761","DOIUrl":"https://doi.org/10.3897/nucet.9.107761","url":null,"abstract":"The paper presents the results of modeling of changes in the isotopic composition of the liquid-salt fuel circulating in the experimental channel of the MBIR reactor facility. The authors tested the ISTAR software environment adapted for solving burnup equations in problems with variable power levels. The loop channel parameters, including two heat exchanger options, were estimated to obtain the appropriate salt transit time through the loop channel zones. Two problems of a circulating fuel system (loop) modeling are considered, namely: (1) modeling the equilibrium salt isotope composition in such a system; and (2) developing a technique for modeling nonstationary isotope kinetics in the MBIR reactor loop. Non-stationary isotope kinetics can be modeled as sequential burnup of nuclides in the neutron field and decay during movement in the external circuit. The authors also developed an algorithm for modeling changes in the isotopic composition of fuel salt during its circulation, taking into account the sequential transfer of a given salt volume from the burnup zone to the zone outside the reactor core. Based on this algorithm, a software package was created using the Python 3.9 programming language and ISTAR modules. In addition, a description of the calculation methodology was given and some calculation results obtained using the software were presented. In the process of working with the program, it was found that, for the given times of the fuel being in each of the zones (2 and 200 seconds, respectively), modeling the change in the isotopic composition during the fuel campaign (500 days) will require the calculation of more than 500 thousand steps. In order to save time, it is necessary to find out whether it will be possible to reduce the number of calls to the neutronic calculation code due to a slight change in the isotopic composition of the fuel in the loop per one burnup step. Work is currently underway to optimize this process.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"77 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82627038","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}
Anatoly A. Kazantsev, Ol’ga V. Supotnitskaya, Evgeniya A. Ivanova, Irina V. Moskovchenko, Ruben I. Mukhamadeev, Vladimir F. Timofeev, Nataliya E. Astakhova
The paper presents the results of simulating a beyond design basis accident with regard to radiolytic hydrogen transport and analysis of hydrogen explosion safety in the reactor cavity and in the central reactor hall of the Bilibino NPP. The KUPOL-M code, version 1.10a, is used as the calculation tool for justifying hydrogen explosion safety. The accident under investigation is a beyond design basis accident, the initial event for which is spontaneous travel of two pairs of automatic control rods and a failure of the reactor scram system. The accident leads to the maximum possible release of positive reactivity, mass destruction of fuel elements, and escape of radiolytic hydrogen, as part of the gas mixture, into the reactor cavity and the central hall and further, through the broken windows, into the atmosphere. The calculation results show that no explosive concentrations of hydrogen are formed in the reactor cavity and in the central hall. Therefore, hydrogen explosion safety is ensured throughout the duration of the design basis accident for the Bilibino NPP unit with the EGP-6 reactor.
{"title":"Hydrogen explosion safety for the Bilibino NPP EGP-6 reactor in conditions of a beyond design basis accident","authors":"Anatoly A. Kazantsev, Ol’ga V. Supotnitskaya, Evgeniya A. Ivanova, Irina V. Moskovchenko, Ruben I. Mukhamadeev, Vladimir F. Timofeev, Nataliya E. Astakhova","doi":"10.3897/nucet.9.107760","DOIUrl":"https://doi.org/10.3897/nucet.9.107760","url":null,"abstract":"The paper presents the results of simulating a beyond design basis accident with regard to radiolytic hydrogen transport and analysis of hydrogen explosion safety in the reactor cavity and in the central reactor hall of the Bilibino NPP. The KUPOL-M code, version 1.10a, is used as the calculation tool for justifying hydrogen explosion safety. The accident under investigation is a beyond design basis accident, the initial event for which is spontaneous travel of two pairs of automatic control rods and a failure of the reactor scram system. The accident leads to the maximum possible release of positive reactivity, mass destruction of fuel elements, and escape of radiolytic hydrogen, as part of the gas mixture, into the reactor cavity and the central hall and further, through the broken windows, into the atmosphere. The calculation results show that no explosive concentrations of hydrogen are formed in the reactor cavity and in the central hall. Therefore, hydrogen explosion safety is ensured throughout the duration of the design basis accident for the Bilibino NPP unit with the EGP-6 reactor.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"5 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135188186","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}
Nikita V. Kovalev, Alexander M. Prokoshin, Alexander S. Kudinov, Vladimir A. Nevinitsa
The VVER-1000 thermal neutron reactor can operate on mixed uranium-plutonium fuel with a content of reactor-grade plutonium up to 5% with a 100% loaded core. In this case, plutonium burns up to 56% of odd isotopes. The energy potential of such plutonium is very low, and its further use in thermal reactors is impractical. However, such plutonium can be used in fast neutron reactors. The paper presents the results of investigating the possibility for such isotopic plutonium composition to be used in the BN-1200 thermal neutron reactor and its value be increased for the plutonium recycle in the reactor. For this purpose, a precision model of the BN-1200 reactor has been developed using the Serpent Monte Carlo code. The model has been verified against the reference values of the nuclear fuel burnup and breeding ratios. The study has shown that such plutonium can be used in the BN-1200 reactor MOX fuel. Maintaining the operating cycle length requires the plutonium fraction in the MOX fuel to be increased up to 2%. In the BN-1200 reactor, the isotopic composition has been found to improve for the further recycle of plutonium in the thermal reactor, i.e. odd plutonium isotopes increase. The fewer odd plutonium isotopes at the beginning of the BN-1200 operating cycle, the greater their increase. It can be seen as the result of the calculation that plutonium from VVER-1000 spent mixed fuel must be loaded into the BN-1200 reactor at least twice to increase the fraction of odd isotopes to the level of reactor-grade plutonium.
{"title":"Use of remix spent mixed fuel plutonium in the BN-1200 reactor","authors":"Nikita V. Kovalev, Alexander M. Prokoshin, Alexander S. Kudinov, Vladimir A. Nevinitsa","doi":"10.3897/nucet.9.107762","DOIUrl":"https://doi.org/10.3897/nucet.9.107762","url":null,"abstract":"The VVER-1000 thermal neutron reactor can operate on mixed uranium-plutonium fuel with a content of reactor-grade plutonium up to 5% with a 100% loaded core. In this case, plutonium burns up to 56% of odd isotopes. The energy potential of such plutonium is very low, and its further use in thermal reactors is impractical. However, such plutonium can be used in fast neutron reactors. The paper presents the results of investigating the possibility for such isotopic plutonium composition to be used in the BN-1200 thermal neutron reactor and its value be increased for the plutonium recycle in the reactor. For this purpose, a precision model of the BN-1200 reactor has been developed using the Serpent Monte Carlo code. The model has been verified against the reference values of the nuclear fuel burnup and breeding ratios. The study has shown that such plutonium can be used in the BN-1200 reactor MOX fuel. Maintaining the operating cycle length requires the plutonium fraction in the MOX fuel to be increased up to 2%. In the BN-1200 reactor, the isotopic composition has been found to improve for the further recycle of plutonium in the thermal reactor, i.e. odd plutonium isotopes increase. The fewer odd plutonium isotopes at the beginning of the BN-1200 operating cycle, the greater their increase. It can be seen as the result of the calculation that plutonium from VVER-1000 spent mixed fuel must be loaded into the BN-1200 reactor at least twice to increase the fraction of odd isotopes to the level of reactor-grade plutonium.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"18 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135188192","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}
Yurii S. Khomyakov, Valerii I. Rachkov, Iurii E. Shvetsov
The lead cooled reactor BREST-OD-300 is developing as a part of Russian federal project “PRORYV”. Two- circuit scheme is used in the reactor for heat removal. An inherent risk of two- circuit reactor is the potential danger of water steam ingression in the core in the case of large leakage in steam generator initiated, for example, Steam Generator Tube Rupture (SGTR). Reactor power and temperature response on vapor penetration to the core is studied, but pressurization effects are not in the purview of the paper. The 3D multi-physics (neutronics + thermal-hydraulics) UNICO-2F code was developed for study of SGTR accident. The code calculates unsteady 3D space dependent distributions of coolant velocity, pressure and temperature, space distributions of vapor concentration and heat release density in the core and 3D temperature distributions in the fuel pins. Guillotine rupture of one tube in Steam Generator (SG) is considered as initial event of the accident. It is shown that even with the most conservative assumptions reactivity insertion due to vapor ingress in the core causes small increase of power in level and as a result maximum cladding temperature continue to stay well below safe operation design limit in the entire transient. Hypothetical option of simultaneous tube rupture in few SG belonging to different loops is also analyzed. It is demonstrated that even in the case of simultaneous large leak in two SG the transient stays mild and temperature in the core after two small oscillations is stabilized at acceptable level. In the long term the analysis confirmed the high level of reactor self-protection against SGTR accident.
{"title":"Coupled modeling of neutronics and thermal-hydraulics processes in LFR under SG-leakage condition","authors":"Yurii S. Khomyakov, Valerii I. Rachkov, Iurii E. Shvetsov","doi":"10.3897/nucet.9.99154","DOIUrl":"https://doi.org/10.3897/nucet.9.99154","url":null,"abstract":"The lead cooled reactor BREST-OD-300 is developing as a part of Russian federal project “PRORYV”. Two- circuit scheme is used in the reactor for heat removal. An inherent risk of two- circuit reactor is the potential danger of water steam ingression in the core in the case of large leakage in steam generator initiated, for example, Steam Generator Tube Rupture (SGTR). Reactor power and temperature response on vapor penetration to the core is studied, but pressurization effects are not in the purview of the paper. The 3D multi-physics (neutronics + thermal-hydraulics) UNICO-2F code was developed for study of SGTR accident. The code calculates unsteady 3D space dependent distributions of coolant velocity, pressure and temperature, space distributions of vapor concentration and heat release density in the core and 3D temperature distributions in the fuel pins. Guillotine rupture of one tube in Steam Generator (SG) is considered as initial event of the accident. It is shown that even with the most conservative assumptions reactivity insertion due to vapor ingress in the core causes small increase of power in level and as a result maximum cladding temperature continue to stay well below safe operation design limit in the entire transient. Hypothetical option of simultaneous tube rupture in few SG belonging to different loops is also analyzed. It is demonstrated that even in the case of simultaneous large leak in two SG the transient stays mild and temperature in the core after two small oscillations is stabilized at acceptable level. In the long term the analysis confirmed the high level of reactor self-protection against SGTR accident.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135188181","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}
Olga V. Nikolaeva, Sergey A. Gaifulin, Leonid P. Bass, Denis V. Dmitriev, Alexandr A. Nikolaev
The paper considers the influence of the spatial grid type on the result of solving the equation of neutron transport in the nuclear power plant (NPP) shielding. Neutron fields were calculated in a realistic model of a liquid metal cooled fast neutron tank reactor with an integral equipment layout. Structured cubic and unstructured hexahedral grids (pmsnsys and FRIGATE codes) and unstructured tetrahedral and prismatic grids (RADUGA T code) are used. Limiting values of the group fluxes averaged over the material zones for refined grids have been obtained. It has been shown that the calculation results depend on the type of approximation for the curvilinear inner boundaries between the material zones rather than on the grid cell type (cube, hexahedron, tetrahedron, prism). Using “toothed” approximations for curvilinear boundaries leads to an increase in the area of the boundaries, as well as to the neutron flux refraction condition arising on them. These effects lead to an upward bias in the transport equation solution, and for all energy groups. Conclusion. When solving an equation of neutron transport in the NPP shielding by a grid technique, it is necessary to use grids other than leading to “toothed” approximations of the inner boundaries. Tetrahedral or prismatic grids, or grids of arbitrary hexahedrons can be recommended, as well as composite grids in which cubic cells are located inside the material zone, and hexahedron cells are located near the zone boundary.
{"title":"Influence of the spatial grid type on the result of calculating the neutron fields in the nuclear power plant shielding","authors":"Olga V. Nikolaeva, Sergey A. Gaifulin, Leonid P. Bass, Denis V. Dmitriev, Alexandr A. Nikolaev","doi":"10.3897/nucet.9.102507","DOIUrl":"https://doi.org/10.3897/nucet.9.102507","url":null,"abstract":"The paper considers the influence of the spatial grid type on the result of solving the equation of neutron transport in the nuclear power plant (NPP) shielding. Neutron fields were calculated in a realistic model of a liquid metal cooled fast neutron tank reactor with an integral equipment layout. Structured cubic and unstructured hexahedral grids (pmsnsys and FRIGATE codes) and unstructured tetrahedral and prismatic grids (RADUGA T code) are used. Limiting values of the group fluxes averaged over the material zones for refined grids have been obtained. It has been shown that the calculation results depend on the type of approximation for the curvilinear inner boundaries between the material zones rather than on the grid cell type (cube, hexahedron, tetrahedron, prism). Using “toothed” approximations for curvilinear boundaries leads to an increase in the area of the boundaries, as well as to the neutron flux refraction condition arising on them. These effects lead to an upward bias in the transport equation solution, and for all energy groups. Conclusion. When solving an equation of neutron transport in the NPP shielding by a grid technique, it is necessary to use grids other than leading to “toothed” approximations of the inner boundaries. Tetrahedral or prismatic grids, or grids of arbitrary hexahedrons can be recommended, as well as composite grids in which cubic cells are located inside the material zone, and hexahedron cells are located near the zone boundary.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135188185","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 heat flux in a Light Water Reactor (LWR) system is used to estimate the Departure from Nucleate Boiling Ratio (DNBR) of the system which is an important thermal hydraulic parameter for nuclear reactors from heat removal point of view. The DNBR signifies an operational safety limit i.e. the nuclear power plant has to be operated with sufficient margin from the specified DNBR limit for assuring its safety. The DNBR is evaluated using a thermal hydraulic analysis code using inputs from neutronics calculation. The present paper presents the evaluation approach of minimum DNBR (MDNBR) during standard neutronics calculation. The DNBR calculation is performed using a core physics analysis code and burnup variation of MDNBR is studied for the full cycle length. The results of calculation are presented using the equilibrium core of 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR). The calculations are performed using VISWAM-TRIHEXFA code system. The few group lattice parametric library for IPWR is generated by lattice analysis code VISWAM. The core follow up calculation for the equilibrium core configuration has been performed using core analysis code TRIHEXFA. A first order thermal hydraulic feedback model has been introduced into the 3D finite difference core simulation tool TRIHEXFA. The critical heat flux calculation, required for estimation of DNBR, has been performed using W-3 Tong and OKB-Gidropress correlations implemented in TRIHEXFA.
{"title":"Evaluation of DNBR with neutronics calculation in LWR systems","authors":"S. Khan, Umasankari Kannan","doi":"10.3897/nucet.9.98452","DOIUrl":"https://doi.org/10.3897/nucet.9.98452","url":null,"abstract":"The heat flux in a Light Water Reactor (LWR) system is used to estimate the Departure from Nucleate Boiling Ratio (DNBR) of the system which is an important thermal hydraulic parameter for nuclear reactors from heat removal point of view. The DNBR signifies an operational safety limit i.e. the nuclear power plant has to be operated with sufficient margin from the specified DNBR limit for assuring its safety. The DNBR is evaluated using a thermal hydraulic analysis code using inputs from neutronics calculation. The present paper presents the evaluation approach of minimum DNBR (MDNBR) during standard neutronics calculation. The DNBR calculation is performed using a core physics analysis code and burnup variation of MDNBR is studied for the full cycle length. The results of calculation are presented using the equilibrium core of 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR). The calculations are performed using VISWAM-TRIHEXFA code system. The few group lattice parametric library for IPWR is generated by lattice analysis code VISWAM. The core follow up calculation for the equilibrium core configuration has been performed using core analysis code TRIHEXFA. A first order thermal hydraulic feedback model has been introduced into the 3D finite difference core simulation tool TRIHEXFA. The critical heat flux calculation, required for estimation of DNBR, has been performed using W-3 Tong and OKB-Gidropress correlations implemented in TRIHEXFA.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"150 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76403491","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 paper deals with issues of designing and creating knowledge bases in the field of nuclear science and technology. The authors present the results of searching for and testing optimal classification and semantic annotation algorithms applied to the textual network content for the convenience of computer-aided filling and updating of scalable semantic repositories (knowledge bases) in the field of nuclear physics and nuclear power engineering and, in the future, for other subject areas, both in Russian and English. The proposed algorithms will provide a methodological and technological basis for creating problem-oriented knowledge bases as artificial intelligence systems, as well as prerequisites for the development of semantic technologies for acquiring new knowledge on the Internet without direct human participation. Testing of the studied machine learning algorithms is carried out by the cross-validation method using corpora of specialized texts. The novelty of the presented study lies in the application of the Pareto optimality principle for multi-criteria evaluation and ranking of the studied algorithms in the absence of a priori information about the comparative significance of the criteria. The project is implemented in accordance with the Semantic Web standards (RDF, OWL, SPARQL, etc.). There are no technological restrictions for integrating the created knowledge bases with third-party data repositories as well as metasearch, library, reference or information and question-answer systems. The proposed software solutions are based on cloud computing using DBaaS and PaaS service models to ensure the scalability of data warehouses and network services. The created software is in the public domain and can be freely replicated.
{"title":"Application of machine learning methods for filling and updating nuclear knowledge bases","authors":"V. Telnov, Y. Korovin","doi":"10.3897/nucet.9.106759","DOIUrl":"https://doi.org/10.3897/nucet.9.106759","url":null,"abstract":"The paper deals with issues of designing and creating knowledge bases in the field of nuclear science and technology. The authors present the results of searching for and testing optimal classification and semantic annotation algorithms applied to the textual network content for the convenience of computer-aided filling and updating of scalable semantic repositories (knowledge bases) in the field of nuclear physics and nuclear power engineering and, in the future, for other subject areas, both in Russian and English. The proposed algorithms will provide a methodological and technological basis for creating problem-oriented knowledge bases as artificial intelligence systems, as well as prerequisites for the development of semantic technologies for acquiring new knowledge on the Internet without direct human participation. Testing of the studied machine learning algorithms is carried out by the cross-validation method using corpora of specialized texts. The novelty of the presented study lies in the application of the Pareto optimality principle for multi-criteria evaluation and ranking of the studied algorithms in the absence of a priori information about the comparative significance of the criteria. The project is implemented in accordance with the Semantic Web standards (RDF, OWL, SPARQL, etc.). There are no technological restrictions for integrating the created knowledge bases with third-party data repositories as well as metasearch, library, reference or information and question-answer systems. The proposed software solutions are based on cloud computing using DBaaS and PaaS service models to ensure the scalability of data warehouses and network services. The created software is in the public domain and can be freely replicated.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"43 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77240492","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}
Evgenia A. Kuleshova, Ivan V. Fedotov, Dmitry A. Maltsev, Aleksey S. Frolov, Nikita V. Stepanov, Denis V. Safonov
Nickel is an essential alloying element in steels used as structural materials in the most common nuclear power reactors of the VVER type. The paper considers reviews the results of structural studies of traditional and advanced materials of the vessels and internals of VVER-type reactors with high nickel contents in their compositions. It is shown that an increased nickel content (up to 5 wt.%) in the steels of VVER pressure vessels contributes to the formation of a more dispersed structure with a smaller size of substructural elements and an increased density of dislocations, as well as a higher volume density of carbide phases. The revealed features of the structure of the reactor pressure vessel steel with high nickel content have the prerequisites for improving the strength and viscoplastic properties due to the increased number of barriers both for the dislocation motion and brittle crack propagation. Using the example of materials for VVER internals, it is shown that the nickel content increased in them up to 25 wt.% contributes to an increase in the volume density of radiation defects (dislocation loops of various types) and radiation-induced phase precipitates (G-phase). As nickel increases from 10 to 25 wt.%, there is a tendency to reduce swelling, which contributes to less shape change of the components of the reactor vessel internals. At the same time, in the steel with the highest nickel content, the highest nickel content was found in the near-boundary regions of the matrix, which contributes to greater austenite stability and a lower probability of the formation of an embrittling α-phase. The data obtained in the work on the effect of nickel alloying on the steel structural phase state and service characteristics were used in the development of new materials for the vessels and internals of advanced reactors.
{"title":"The role of nickel in forming a structure providing increased service properties of reactor structural materials","authors":"Evgenia A. Kuleshova, Ivan V. Fedotov, Dmitry A. Maltsev, Aleksey S. Frolov, Nikita V. Stepanov, Denis V. Safonov","doi":"10.3897/nucet.9.102914","DOIUrl":"https://doi.org/10.3897/nucet.9.102914","url":null,"abstract":"Nickel is an essential alloying element in steels used as structural materials in the most common nuclear power reactors of the VVER type. The paper considers reviews the results of structural studies of traditional and advanced materials of the vessels and internals of VVER-type reactors with high nickel contents in their compositions. It is shown that an increased nickel content (up to 5 wt.%) in the steels of VVER pressure vessels contributes to the formation of a more dispersed structure with a smaller size of substructural elements and an increased density of dislocations, as well as a higher volume density of carbide phases. The revealed features of the structure of the reactor pressure vessel steel with high nickel content have the prerequisites for improving the strength and viscoplastic properties due to the increased number of barriers both for the dislocation motion and brittle crack propagation. Using the example of materials for VVER internals, it is shown that the nickel content increased in them up to 25 wt.% contributes to an increase in the volume density of radiation defects (dislocation loops of various types) and radiation-induced phase precipitates (G-phase). As nickel increases from 10 to 25 wt.%, there is a tendency to reduce swelling, which contributes to less shape change of the components of the reactor vessel internals. At the same time, in the steel with the highest nickel content, the highest nickel content was found in the near-boundary regions of the matrix, which contributes to greater austenite stability and a lower probability of the formation of an embrittling α-phase. The data obtained in the work on the effect of nickel alloying on the steel structural phase state and service characteristics were used in the development of new materials for the vessels and internals of advanced reactors.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"28 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135186994","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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