High-Order Homogenization Method in Diffusion Theory for Reactor Core Simulation and Design Calculation

Most modern nodal methods in use by the reactor vendors and utilities are based on the generalized equivalence theory (GET) that uses homogenized cross sections and flux discontinuity factors. These homogenized parameters, referred to as infinite medium parameters, are precomputed by performing single bundle fine-mesh calculations with zero current boundary conditions. It is known that for configurations in which the node-to-node leakage (e.g., surface current-to-flux ratio) is large the use of the infinite medium parameters could lead to large errors in the nodal solution. This would be the case for highly heterogeneous core configurations, typical of modern reactor core designs

An innovative three-dimensional heterogeneous coarse-mesh transport method for advanced and generation IV reactor core analysis and design

Issued as final reportUnited States. Department of Energ

Consistent Multigroup Theory Enabling Accurate Course-Group Simulation of Gen IV Reactors

The objective of this proposal is the development of a consistent multi-group theory that accurately accounts for the energy-angle coupling associated with collapsed-group cross sections. This will allow for coarse-group transport and diffusion theory calculations that exhibit continuous energy accuracy and implicitly treat cross- section resonances. This is of particular importance when considering the highly heterogeneous and optically thin reactor designs within the Next Generation Nuclear Plant (NGNP) framework. In such reactors, ignoring the influence of anisotropy in the angular flux on the collapsed cross section, especially at the interface between core and reflector near which control rods are located, results in inaccurate estimates of the rod worth, a serious safety concern. The scope of this project will include the development and verification of a new multi-group theory enabling high-fidelity transport and diffusion calculations in coarse groups, as well as a methodology for the implementation of this method in existing codes. This will allow for a higher accuracy solution of reactor problems while using fewer groups and will reduce the computational expense. The proposed research represents a fundamental advancement in the understanding and improvement of multi- group theory for reactor analysis

An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulator

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis

Effects of Spatial Variations in Packing Fraction on Reactor Physics Parameters in Pebble-Bed Reactors

The well-known spatial variation of packing fraction near the outer boundary of a pebble-bed reactor core is cited. The ramifications of this variation are explored with the MCNP computer code. It is found that the variation has negligible effects on the global reactor physics parameters extracted from the MCNP calculations for use in analysis by diffusion-theory codes, but for local reaction rates the effects of the variation are naturally important. Included is some preliminary work in using first-order perturbation theory for estimating the effect of the spatial variation of packing fraction on the core eigenvalue and the fision density distribution

FHR Materials PIRT Report

Phenomena Identification and Ranking Table (PIRT) report for metallic materials use in Fluoride High-Temperature Reactors (FHRs

Hybrid coarse mesh transport/diffusion method for PBR neutronics analysis

Issued as final reportIdaho National Engineering Laborator

NRC fellowships in the nuclear and radiological engineering program at Georgia Tech

Issued as final reportU.S. Nuclear Regulatory Commissio

An innovative Monte Carlo adaptation of a heterogeneous coarse mesh transport method for efficient and accurate radiation dose predicition in breast and prostrate cancer patients

Issued as final reportGeorgia Cancer Coalitio