Abstract

Due to its low computational cost, nodal diffusion methods are still commonly used to simulate a full core reactor model. Even though many approximation procedures are involved in the diffusion equation, the resulted accuracy is still worth the computational cost involved. With the correct treatment, an excellent comparison with the high resolution deterministic or Monte Carlo lattice codes can be obtained. This work represents the developmental effort to build an accurate nodal kernel to treat hexagonal geometry in PARCS. Developed originally for Light Water Reactors (LWRs), this U.S. N.R.C nodal diffusion code has been enhanced to handle many advanced reactors, including ones that involve non-cartesian geometries. Hexagonal assembly reactors are not something new. However, the complexity of this type of geometry still provides challenges to accurately model in nodal diffusion methods. An alternative to solve hexagonal geometry problems is to explicitly treat each hexagonal node into six triangular nodes. A method called Triangular Polynomial Expansion Nodal (TriPEN) has been developed using this approach. For fast spectrum reactors, a 4-term quadratic polynomial expansion for the flux is used. On the other hand, the shorter mean free path in thermal spectrum reactors requires higher-order polynomials, so a 9-term cubic polynomial series is utilized. To validate the method performance, sodium-cooled fast reactor models and VVER models were analyzed using the TriPEN method and the eigenvalue and power distribution were compared with the high-fidelity calculation from the Serpent Monte Carlo code.

Bio

Dr. Yunlin Xu had received his bachelor’s degree in 1990 and master’s degree in 1991, both from Tsinghua University, and received his Ph. D from Purdue in 2004, all majored in Nuclear Engineering. He had also received a master’s degree in computer science from Purdue in 2006.

He was associate professor at Tsinghua University from 1991 to 1999. his research was in the area of nuclear reactor physics design and safety analysis. He was responsible for the reactor physics design of two advanced nuclear reactors: a 200 MW nuclear heating reactor (NHR) and a 10 MW high temperature gas-cooled reactor. After he come to Purdue university at end of 1999, his research area has switched to methods and code development for Neutron kinetics and thermal-fluid coupling simulation, multi-physics coupling.  At 2008, Dr Xu left Purdue to join Argonne National Laboratory but worked off campus at University of Michigan. 10 years later, Dr. Xu come back to Purdue as a faculty member of School of nuclear engineering. His research interest also includes large linear and non-linear system solution methods. He was primary developer of neutronics codes such as PARCS, DeCart, MPACT. He also contributes to nuclear reactor system codes such as TRACE, RELAP. His researches are shown in more than 100 published papers.