Convergence Analysis of the Nonlinear Coarse-Mesh Finite Difference Method for One-Dimensional Fixed-Source Neutron Diffusion Problem

Lee, D.,Downar, T. J.,Kim, Y. AMERICAN NUCLEAR SOCIETY 2004 Nuclear science and engineering Vol.147 No.2

Multi-group SP₃ approximation for simulation of a three-dimensional PWR rod ejection accident

Lee, Deokjung,Kozlowski, Tomasz,Downar, Thomas J. Elsevier 2015 Annals of nuclear energy Vol.77 No.-

<P><B>Abstract</B></P> <P>Previous researchers have shown that the simplified P<SUB>3</SUB> (SP<SUB>3</SUB>) approximation is capable of providing sufficiently high accuracy for both static and transient simulations for reactor core analysis with considerably less computational expense than higher order transport methods such as the discrete ordinate or the full spherical harmonics methods. The objective of this paper is to provide a consistent comparison of two-group (2G) and multi-group (MG) diffusion and SP<SUB>3</SUB> transport for rod ejection accident (REA) in a practical light water reactor (LWR) problem. The analysis is performed on two numerical benchmarks, a 3×3 assembly mini-core and a full pressurized water reactor (PWR) core. The calculations were performed using pin homogenized and assembly homogenized cross sections for a series of benchmarks of increasing difficulty, in two-dimensional (2D) and three-dimensional (3D), 2G and MG, diffusion and transport, as well as with and without feedback. All results show consistency with the reference results obtained from higher-order methods. It is demonstrated that the analyzed problems show small group-homogenization effects, but relatively significant transport effects which are satisfactorily addressed by the SP<SUB>3</SUB> transport method. The sensitivity tests also show that, for the REA simulation, the MG is more conservative than 2G, P<SUB>1</SUB> is more conservative than SP<SUB>3</SUB> for a 1/3 MOX loaded full-core problem.</P> <P><B>Highlights</B></P> <P> <UL> <LI> The multi-group SP<SUB>3</SUB> method developed and implemented in PARCS for the MOX analysis. </LI> <LI> The verifications were performed in 2D and 3D, 2G and MG, diffusion and transport, with and without feedback. </LI> <LI> All results show consistency with the reference results obtained from the ANL P<SUB>N</SUB> transport code VARIANT for steady-state and transport calculations. </LI> <LI> It was found that the SP<SUB>3</SUB> angular approximation captures sufficient transport effects for both steady-state and transient, and provides essentially the same results as the VARIANT P<SUB>5</SUB> method. </LI> <LI> From the transient results of the full-core problem, it was noted that MG is more conservative than 2G, and P<SUB>1</SUB> is more conservative than SP<SUB>3</SUB>. </LI> </UL> </P>

Transient simulation of an endothermic chemical process facility coupled to a high temperature reactor: Model development and validation

Brown, N.R.,Seker, V.,Revankar, S.T.,Downar, T.J. North-Holland Pub. Co 2012 Nuclear engineering and design Vol.248 No.-

A high temperature reactor (HTR) is a candidate to drive high temperature water-splitting using process heat. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Three thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. The models and coupling scheme are presented here, as well as a transient test case initiated within the chemical plant. The 50% feed flow failure within the chemical plant results in a slow loss-of-heat sink (LOHS) accident in the nuclear reactor. Due to the temperature feedback within the reactor core the nuclear reactor partially shuts down over 1500s. Two distinct regions are identified within the coupled plant response: (1) immediate LOHS due to the loss of the sulfuric acid decomposition section and (2) continuing slow LOHS due to the chemical species cascade throughout the plant.

Fourier Convergence Analysis of Two-Dimensional/One-Dimensional Coupling Methods for the Three-Dimensional Neutron Diffusion Eigenvalue Problems

Lee, H. C.,Noh, J. M.,Joo, H. K.,Lee, D.,Downar, T. J. Academic Press 2007 Nuclear science and engineering Vol.156 No.1

Validation of Numerical Methods to Calculate Bypass Flow in a Prismatic Gas-Cooled Reactor Core

탁남일,김민환,임홍식,노재만,Timothy J. Drzewiecki,Volkan Seker,Thomas J. Downar,JOSEPH KELLY 한국원자력학회 2013 Nuclear Engineering and Technology Vol.45 No.6

For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR), intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI) and the AGREE code of the University of Michigan (U of M). One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU) in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

Cross section generation for a conceptual horizontal, compact high temperature gas reactor

Kang Junsu,Seker Volkan,Ward Andrew,Jabaay Daniel,Kochunas Brendan,Downar Thomas 한국원자력학회 2024 Nuclear Engineering and Technology Vol.56 No.3

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

Tak, Nam-Il,Kim, Min-Hwan,Lim, Hong-Sik,Noh, Jae Man,Drzewiecki, Timothy J.,Seker, Volkan,Downar, Thomas J.,Kelly, Joseph Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.6

For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR), intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI) and the AGREE code of the University of Michigan (U of M). One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU) in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.