Ratio correction, temperature feedback, and other advances for the fission-matrix based hybrid transport code RAPID

Open Access
- Author:
- He, Donghao
- Graduate Program:
- Nuclear Engineering
- Degree:
- Doctor of Philosophy
- Document Type:
- Dissertation
- Date of Defense:
- February 25, 2020
- Committee Members:
- William J Walters, Dissertation Advisor/Co-Advisor
William J Walters, Committee Chair/Co-Chair
Leigh Winfrey, Committee Member
Azaree Lintereur, Committee Member
Jerry Y Harrington, Outside Member
Arthur Thompson Motta, Program Head/Chair - Keywords:
- RAPID
Fission Matrix
Neutron Transport
Temperature Feedback
Uncertainty Analysis - Abstract:
- A novel hybrid deterministic-stochastic neutron transport code RAPID has been previously developed. It features high-accuracy neutron transport calculation like Monte Carlo codes, while preserving the fast calculation speed of deterministic codes. RAPID is based on the fission matrix theory, and implements the fission matrix combination technique to quickly estimate the system fission matrix. RAPID has been proven accurate on spent fuel system or homogeneous reactor core loading, but it suffers from errors in the heterogeneous core loading. The goal of this work is to improve the RAPID code by reducing the material discontinuity error; analyzing the numerical solution methods including fission-matrix homogenization; developing temperature-feedback calculation capability; and providing uncertainty analysis of RAPID. In order to reduce the error because of material discontinuity, a correction ratio method has been developed for RAPID and applied on the BEAVRS benchmark model at hot- zero-power condition. The correction ratio greatly reduces the error, and RAPID provides a high-fidelity 3D whole-core pin-wise fission distribution, compared to a Serpent 2 Monte Carlo reference calculation. In addition, the correction ratio is extended to multiple axial slices to account for the partially inserted control rod effect. RAPID accurately estimates the differential and integral control rod worth of different banks in BEAVRS. The transport calculation results, including keff, fission source distribution and control rod worth, are compared between RAPID and a neutron diffusion code PARCS. It shows that RAPID provides better accuracy calculations with only a relatively small increase in computational costs compared to higher-order transport methods. Fission-matrix homogenization, together with other numerical methods in RAPID are also investigated in this work. The fission matrix homogenization with pin power reconstruction, which had previously been implemented but not analyzed, is demonstrated to greatly accelerate the RAPID calculation with only minimal loss of accuracy. RAPID is also extended to model fuel and moderator temperature distributions in the BEAVRS model. Two techniques are used to interpolate the fission matrix database according to the local fuel and moderator temperature. It proves that the interpolation techniques allow RAPID to perform accurate transport calculations with heterogeneous temperature distribution. Finally, two fission matrix re-samplings technique are applied in RAPID to analyze the uncertainty in calculation results due to fission-matrix uncertainty.