Volkan Seker - Academia.edu (original) (raw)
Papers by Volkan Seker
Transactions of the American Nuclear Society, 2014
In order to improve the convergence of the AGREE code for porous medium, a Newton-Krylov solver w... more In order to improve the convergence of the AGREE code for porous medium, a Newton-Krylov solver was developed for steady state problems. The current three-equation system was expanded and then coupled using Newton's Method. Theoretical behavior predicts second order convergence, while actual behavior was highly nonlinear. The discontinuous derivatives found in both closure and empirical relationships prevented true second order convergence. Agreement between the current solution and new Exact Newton solution was well below the convergence criteria. While convergence time did not dramatically decrease, the required number of outer iterations was reduced by approximately an order of magnitude. GMRES was also used to solve problem, where ILU without fill-in was used to precondition the iterative solver, and the performance was slightly slower than the direct solution. (authors)
Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts... more Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology roadmap. Considerable research has been performed on the design and safety analysis of these reactors. However, the codes and methods being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun’s resistance model for pebble bed and energy equation in two temperature model are solved in three-dimensional cylindrical geometry. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. Various experimental cases are modeled and good agreement is observed. The...
Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts... more Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology roadmap. Considerable research has been performed on the design and safety analysis of these reactors. However, the codes and methods being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun’s resistance model for pebble bed and energy equation in two temperature model are solved in three-dimensional cylindrical geometry. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. Various experimental cases are modeled and good agreement is observed. The...
Transactions of the American Nuclear Society, 2007
The OECD/NEA/NSC PBMR-400 benchmark problem was developed to support the validation and verificat... more The OECD/NEA/NSC PBMR-400 benchmark problem was developed to support the validation and verification efforts for the PBMR design. This paper describes the analysis of this problem using the PARCS-DIREKT coupled code system. The benchmark problem involved the use of two different cross-section libraries, one which was generated from a VSOP equilibrium core calculation and has no dependence on core conditions. The second library provides for dependence on five state parameters and was designed for transient analysis. The paper here reports the steady-state cases using the VSOP set of cross-sections. The results are shown to be in good agreement with those of VSOP. Also reported here are the results of the steady-state thermal-hydraulic DIRECKT solution with a given power profile obtained from VSOP equilibrium core calculation. This analysis provides some insight as to the most important parameters in the design of PBMR-400. (authors)
Transactions of the American Nuclear Society, 2007
Transactions of the American Nuclear Society, 2011
The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts ... more The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivated the development of more accurate and efficient tools for the design and safety evaluations of the PMR. In addition to the work invested in new methods, it is essential to develop appropriate benchmarks to verify and validate the new methods in computer codes. The purpose of this benchmark is to establish a well-defined problem, based on a common given set of data, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events. The be...
A core fluids and heat transfer model has been developed for the prismatic high temperature gas r... more A core fluids and heat transfer model has been developed for the prismatic high temperature gas reactor in support of the US NRC Next Generation Nuclear Plant (NGNP) evaluation model. The core fluids modeling relies on a subchannel approach in which the primary coolant flow path through the core region and vertical in-core and ex-core gaps can be modeled as individual subchannels. These subchannels are connected together to represent a three dimensional reactor. An initial validation calculation for the core fluids model has been performed using data available in literature for bypass flow. The predicted bypass flow was within 2.6% of the value reported in the literature. The core level heat transfer model is based on a triangular finite volume method, where the base triangle is one sixth of the prismatic block. In order to improve the spatial accuracy at this level, a triangular refinement method was also implemented. The fuel compact temperature is calculated by a cylindrical cond...
A computational code system based on coupling the Monte Carlo code MCNP5 and the Computational Fl... more A computational code system based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program “McSTAR”. McSTAR is written in FORTRAN90 programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STARCD for every region. Three different methods were investigated and implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping ...
The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water r... more The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water reactor (LWR) homogenization methods rely on the short mean free paths of neutrons in LWR. In gas-cooled, graphite-moderated reactors like the HTTR neutrons have much longer mean free paths and penetrate further into neighboring assemblies than in LWRs. Because of this, conventional lattice calculations with a single assembly may not be valid. In addition to difficulties caused by the longer mean free paths, the HTTR presents unique axial and radial heterogeneities that require additional modifications to the single assembly homogenization method. To handle these challenges, the homogenization domain is decreased while the computational domain is increased. Instead of homogenizing a single hexagonal fuel assembly, the assembly is split into six triangles on the radial plane and five blocks axially in order to account for the placement of burnable poisons. Furthermore, the radial domain is...
The ability to accurately predict local pin powers in nuclear reactors is necessary to understand... more The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques based on diffusion theory and pin power reconstruction (PPR). The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is 'post-refinement' and thus has no impact on the global solution. (authors)
This paper presents the uncertainty analysis of the original Minimum Critical Core of the TREAT r... more This paper presents the uncertainty analysis of the original Minimum Critical Core of the TREAT reactor. A stochastic uncertainty analysis was performed on the contribution of geometric and composition specification of the assembly using Monte Carlo Neutronics methods. Results show that Boron contamination, Zr can thickness, Al can thickness and flat-to-flat distance of fuel blocks are the most significant design factors contributing to the variance.
Journal of Physics: Conference Series
The High Temperature Test Reactor is a 30-MWth helium-cooled, graphite-moderated, prismatic-type ... more The High Temperature Test Reactor is a 30-MWth helium-cooled, graphite-moderated, prismatic-type gas reactor developed by the Japan Atomic Energy Agency (JAEA). The hexagonal shape of fuel blocks in the HTTR core combined with complex inner structures containing TRISO particles results in the double-heterogeneity effect that increases the simulation challenge of the reactor. This research has a goal to accurately model the HTTR fuel blocks employing the standard two-step procedure of a reactor analysis: employ a lattice physics calculation to generate homogeneous cross sections and use them in a nodal diffusion calculation. The implementation of the diffusion approximation results in a faster calculation with acceptable accuracy compared to the high-resolution lattice calculation. An advanced method called Triangular Polynomial Expansion Nodal (TriPEN) method was used in this work for the nodal diffusion calculation to accurately model the flux discontinuity effect between blocks by...
Nuclear Technology
Abstract The paper presents a self-adaptive feature selection algorithm we developed for solving ... more Abstract The paper presents a self-adaptive feature selection algorithm we developed for solving high-dimensional uncertainty quantification problems. The development of the algorithm was motivated and supported by the benchmarking of the Transient Reactor Test (TREAT) transient test 2857. The generalized polynomial chaos expansion scheme was adopted to decompose the response functions. Our algorithm was applied to select the dominant basis from the candidate polynomial basis in a self-adaptive manner by assigning weights to the polynomial basis and adjusting the weights using the least absolute shrinkage and selection operator regularization–estimated coefficients through iterations. The developed algorithm can recognize the significant basis terms in the polynomial expansion of the response functions and therefore build a sparse polynomial expansion using a limited number of samples. The algorithm was implemented and verified through three different TREAT modeling cases. The testing results demonstrated the general stability and prediction performance of our algorithm and provided useful information about the uncertainty mechanism of the TREAT transient test 2857.
Transactions of the American Nuclear Society, 2014
In order to improve the convergence of the AGREE code for porous medium, a Newton-Krylov solver w... more In order to improve the convergence of the AGREE code for porous medium, a Newton-Krylov solver was developed for steady state problems. The current three-equation system was expanded and then coupled using Newton's Method. Theoretical behavior predicts second order convergence, while actual behavior was highly nonlinear. The discontinuous derivatives found in both closure and empirical relationships prevented true second order convergence. Agreement between the current solution and new Exact Newton solution was well below the convergence criteria. While convergence time did not dramatically decrease, the required number of outer iterations was reduced by approximately an order of magnitude. GMRES was also used to solve problem, where ILU without fill-in was used to precondition the iterative solver, and the performance was slightly slower than the direct solution. (authors)
Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts... more Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology roadmap. Considerable research has been performed on the design and safety analysis of these reactors. However, the codes and methods being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun’s resistance model for pebble bed and energy equation in two temperature model are solved in three-dimensional cylindrical geometry. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. Various experimental cases are modeled and good agreement is observed. The...
Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts... more Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology roadmap. Considerable research has been performed on the design and safety analysis of these reactors. However, the codes and methods being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun’s resistance model for pebble bed and energy equation in two temperature model are solved in three-dimensional cylindrical geometry. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. Various experimental cases are modeled and good agreement is observed. The...
Transactions of the American Nuclear Society, 2007
The OECD/NEA/NSC PBMR-400 benchmark problem was developed to support the validation and verificat... more The OECD/NEA/NSC PBMR-400 benchmark problem was developed to support the validation and verification efforts for the PBMR design. This paper describes the analysis of this problem using the PARCS-DIREKT coupled code system. The benchmark problem involved the use of two different cross-section libraries, one which was generated from a VSOP equilibrium core calculation and has no dependence on core conditions. The second library provides for dependence on five state parameters and was designed for transient analysis. The paper here reports the steady-state cases using the VSOP set of cross-sections. The results are shown to be in good agreement with those of VSOP. Also reported here are the results of the steady-state thermal-hydraulic DIRECKT solution with a given power profile obtained from VSOP equilibrium core calculation. This analysis provides some insight as to the most important parameters in the design of PBMR-400. (authors)
Transactions of the American Nuclear Society, 2007
Transactions of the American Nuclear Society, 2011
The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts ... more The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivated the development of more accurate and efficient tools for the design and safety evaluations of the PMR. In addition to the work invested in new methods, it is essential to develop appropriate benchmarks to verify and validate the new methods in computer codes. The purpose of this benchmark is to establish a well-defined problem, based on a common given set of data, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events. The be...
A core fluids and heat transfer model has been developed for the prismatic high temperature gas r... more A core fluids and heat transfer model has been developed for the prismatic high temperature gas reactor in support of the US NRC Next Generation Nuclear Plant (NGNP) evaluation model. The core fluids modeling relies on a subchannel approach in which the primary coolant flow path through the core region and vertical in-core and ex-core gaps can be modeled as individual subchannels. These subchannels are connected together to represent a three dimensional reactor. An initial validation calculation for the core fluids model has been performed using data available in literature for bypass flow. The predicted bypass flow was within 2.6% of the value reported in the literature. The core level heat transfer model is based on a triangular finite volume method, where the base triangle is one sixth of the prismatic block. In order to improve the spatial accuracy at this level, a triangular refinement method was also implemented. The fuel compact temperature is calculated by a cylindrical cond...
A computational code system based on coupling the Monte Carlo code MCNP5 and the Computational Fl... more A computational code system based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program “McSTAR”. McSTAR is written in FORTRAN90 programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STARCD for every region. Three different methods were investigated and implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping ...
The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water r... more The High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditional light water reactor (LWR) homogenization methods rely on the short mean free paths of neutrons in LWR. In gas-cooled, graphite-moderated reactors like the HTTR neutrons have much longer mean free paths and penetrate further into neighboring assemblies than in LWRs. Because of this, conventional lattice calculations with a single assembly may not be valid. In addition to difficulties caused by the longer mean free paths, the HTTR presents unique axial and radial heterogeneities that require additional modifications to the single assembly homogenization method. To handle these challenges, the homogenization domain is decreased while the computational domain is increased. Instead of homogenizing a single hexagonal fuel assembly, the assembly is split into six triangles on the radial plane and five blocks axially in order to account for the placement of burnable poisons. Furthermore, the radial domain is...
The ability to accurately predict local pin powers in nuclear reactors is necessary to understand... more The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques based on diffusion theory and pin power reconstruction (PPR). The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is 'post-refinement' and thus has no impact on the global solution. (authors)
This paper presents the uncertainty analysis of the original Minimum Critical Core of the TREAT r... more This paper presents the uncertainty analysis of the original Minimum Critical Core of the TREAT reactor. A stochastic uncertainty analysis was performed on the contribution of geometric and composition specification of the assembly using Monte Carlo Neutronics methods. Results show that Boron contamination, Zr can thickness, Al can thickness and flat-to-flat distance of fuel blocks are the most significant design factors contributing to the variance.
Journal of Physics: Conference Series
The High Temperature Test Reactor is a 30-MWth helium-cooled, graphite-moderated, prismatic-type ... more The High Temperature Test Reactor is a 30-MWth helium-cooled, graphite-moderated, prismatic-type gas reactor developed by the Japan Atomic Energy Agency (JAEA). The hexagonal shape of fuel blocks in the HTTR core combined with complex inner structures containing TRISO particles results in the double-heterogeneity effect that increases the simulation challenge of the reactor. This research has a goal to accurately model the HTTR fuel blocks employing the standard two-step procedure of a reactor analysis: employ a lattice physics calculation to generate homogeneous cross sections and use them in a nodal diffusion calculation. The implementation of the diffusion approximation results in a faster calculation with acceptable accuracy compared to the high-resolution lattice calculation. An advanced method called Triangular Polynomial Expansion Nodal (TriPEN) method was used in this work for the nodal diffusion calculation to accurately model the flux discontinuity effect between blocks by...
Nuclear Technology
Abstract The paper presents a self-adaptive feature selection algorithm we developed for solving ... more Abstract The paper presents a self-adaptive feature selection algorithm we developed for solving high-dimensional uncertainty quantification problems. The development of the algorithm was motivated and supported by the benchmarking of the Transient Reactor Test (TREAT) transient test 2857. The generalized polynomial chaos expansion scheme was adopted to decompose the response functions. Our algorithm was applied to select the dominant basis from the candidate polynomial basis in a self-adaptive manner by assigning weights to the polynomial basis and adjusting the weights using the least absolute shrinkage and selection operator regularization–estimated coefficients through iterations. The developed algorithm can recognize the significant basis terms in the polynomial expansion of the response functions and therefore build a sparse polynomial expansion using a limited number of samples. The algorithm was implemented and verified through three different TREAT modeling cases. The testing results demonstrated the general stability and prediction performance of our algorithm and provided useful information about the uncertainty mechanism of the TREAT transient test 2857.