Abstract
Nuclear reactor systems that use fuel dissolved in a liquid have the potential for enhanced safety characteristics, improved fuel-cycle outcomes, and more efficient isotope-production configurations. In these reactor systems, the fueled liquid may simultaneously undergo irradiation, physical and chemical removal processes, and fueling. The modeling and simulation of this transmutation and decay with material additions and removals is an ongoing research area. An accurate simulation tool is critical to the reactor and fuel-cycle design, reactor deployment, and source-term characterization for these advanced reactor systems. The work described herein involved implementing, testing, and applying the capability to perform reactor physics simulations within the Oak Ridge National Laboratory-developed SCALE suite for nuclear systems analyses and design, leveraging much of its pedigree in quality-assurance and reactor-analysis capabilities. The functionalities to simulate irradiation with material feeds and removals had been added in ORIGEN, and the TRITON reactor physics sequence was extended to calculate the total removed material and track external nonirradiated mixtures to estimate separate processing or waste streams. Results from these capabilities align with analytical expectations obtained from ORIGEN for simplified test cases and with expectations for a molten salt reactor application. This implementation, available with the SCALE 6.3 release, provides for a more efficient and accurate material accountability methodology, allowing for the characterization, design, and analysis of the complete isotopic material inventory of advanced liquid-fueled systems for a variety of applications.
| Original language | English |
|---|---|
| Article number | 110236 |
| Journal | Annals of Nuclear Energy |
| Volume | 196 |
| DOIs | |
| State | Published - Feb 2024 |
Funding
This work was supported by numerous sponsors, including the US Department of Energy , Office of Technology Transitions , Technology Commercialization Fund and the US Nuclear Regulatory Commission Office of Research . In particular, the authors would like to thank Jeff Powers for his initial prototype scripts, which first became a revised set of scripts in “ChemTriton” and were later integrated into SCALE/TRITON as presented in this paper. The authors would also like to thank Austin Lo for his testing contributions during his postdoctoral research time at Oak Ridge National Laboratory and Kevin Clarno at University of Texas, Austin, for his early testing and feedback based on SCALE 6.3 beta releases. This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan).This work was supported by numerous sponsors, including the US Department of Energy, Office of Technology Transitions, Technology Commercialization Fund and the US Nuclear Regulatory Commission Office of Research. In particular, the authors would like to thank Jeff Powers for his initial prototype scripts, which first became a revised set of scripts in “ChemTriton” and were later integrated into SCALE/TRITON as presented in this paper. The authors would also like to thank Austin Lo for his testing contributions during his postdoctoral research time at Oak Ridge National Laboratory and Kevin Clarno at University of Texas, Austin, for his early testing and feedback based on SCALE 6.3 beta releases.
Keywords
- Continuous feed
- Depletion
- Fuel cycle
- MSRE
- Molten salt reactor
- Nuclide removal
- SCALE