Abstract
Fuel fragmentation, relocation, and dispersal are some of the largest issues remaining in the nuclear industry before rod-average burnup can be increased beyond 62 GWd/tU. The issue is primarily related to the potential for fuel to be dispersed into the reactor primary system, which may increase public risk. One way to prevent dispersal is to avoid cladding burst. The objective of this work is to support the high burnup safety case by evaluating cladding burst under high-burnup, full-length fuel rods and to identify uncertainties that could improve model predictions. The results of this analysis will evaluate realistic, prototypic loss-of-coolant accident (LOCA) conditions; support future cladding burst test designs; and inform the development of mechanistic material models. Realistic high-burnup operating conditions were implemented in the BISON fuel performance code to simulate steady-state and LOCA transient fuel rod evolution to the point at which cladding burst occurred. Parametric studies are performed to assess code response to changes in rod internal pressures and heating rates. Results were compared with simulated LOCA experiments to identify inconsistencies between commercial fuel rod analysis and experimental validation. The representative full-length fuel rod LOCA simulation results did not agree with cladding burst tests. Cladding burst tests indicated burst occurring well below (100–150 °C) those calculated in the full-length fuel rod LOCA analysis. Further investigation indicated that the cladding burst tests do not appear to be representative for full-length fuel rods. The inconsistency investigated in this work showed that the differences between the BISON simulation and the experiment's cladding burst conditions arise from an incomplete characterization of the cladding surface temperature, detailed rodlet characterization, lack of cladding strain measurements, and uncertainty in the cladding creep and failure models.
Original language | English |
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Article number | 153621 |
Journal | Journal of Nuclear Materials |
Volume | 563 |
DOIs | |
State | Published - May 2022 |
Funding
The authors would like to thank the Nuclear Energy Advanced Modeling and Simulation program of the US Department of Energy's (DOE's) Office of Nuclear Energy for supporting the fuel analysis. Furthermore, the authors would like to express appreciation to Jason Harp and Caleb Massey of the ORNL Nuclear Energy and Fuel Cycle Division for their support in the review of this manuscript. Lastly, this research used the resources of the High Performance Computing Center at Idaho National Laboratory, which is supported by DOE ’s Office of Nuclear Energy and the Nuclear Science User Facilities under contract no. DE-AC07-05ID14517. This manuscript has been authored by UT-Battelle, LLC under Contract No. 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 ). The authors would like to thank the Nuclear Energy Advanced Modeling and Simulation program of the US Department of Energy's (DOE's) Office of Nuclear Energy for supporting the fuel analysis. Furthermore, the authors would like to express appreciation to Jason Harp and Caleb Massey of the ORNL Nuclear Energy and Fuel Cycle Division for their support in the review of this manuscript. Lastly, this research used the resources of the High Performance Computing Center at Idaho National Laboratory, which is supported by DOE ?s Office of Nuclear Energy and the Nuclear Science User Facilities under contract no. DE-AC07-05ID14517.