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A study on the impact of using a subchannel resolution for modeling of large break loss of coolant accidents

Journal Article · · Annals of Nuclear Energy
 [1];  [1];  [1];  [1]
  1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
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The nuclear industry is investigating the feasibility of transitioning from 18- to 24-month fuel cycles because of the positive impact it would have on the operational costs for the current fleet of light-water reactors. A challenge to making this change is the increased risk of fuel fragmentation, relocation, and dispersal (FFRD) due to the known potential for ceramic fuel to pulverize into fine particles at the higher discharge burnups. Previous work has been performed by the Nuclear Energy Advanced Modeling and Simulation program to assess FFRD risk in high-burnup cores using the BISON fuel performance code and a coarse mesh thermal hydraulics (T/H) solution for a loss-of-coolant accident (LOCA) using the TRACE system T/H code. Because of the importance of the T/H solution for FFRD assessment, this study seeks to investigate the impact of using higher-fidelity subchannel techniques for modeling of the LOCA transient. CTF was used to model a subregion of a high-burnup core that was depleted by the Virtual Environment for Reactor Applications (VERA) multiphysics core simulator. Both coarse-mesh and pin-resolved models were created in CTF, and a consistent coarse-mesh TRACE model was also developed to allow for benchmarking the code results. Further, a large-break loss-of-coolant accident (LBLOCA) reflood transient was simulated using these three models, and results were compared. Results showed some consistent differences between the CTF and TRACE coarse models, including a higher peak cladding temperature (PCT) prediction in CTF and later quenching in CTF; however, the transient clad temperature behavior was similar, and these differences are likely due to post-critical heat flux heat transfer modeling differences and minimum film boiling temperature model differences. The pin-resolved results indicate that the PCT in the lumped model is often under-predicted by as much as 70 °C and that PCT occurs at a different location than the high-power pin in the assembly. The lumped model predicts a difference of 10 °C or less between the average and hot pins in the assembly, whereas the pin-resolved model predicts a range of over 100 °C. These results indicate that higher-fidelity T/H results may have an impact on predicted core behavior during LOCA, which may be important to consider when assessing FFRD risk.
Research Organization:
Idaho National Laboratory (INL), Idaho Falls, ID (United States); Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE)
Grant/Contract Number:
AC05-00OR22725; AC07-05ID14517
OSTI ID:
2397455
Journal Information:
Annals of Nuclear Energy, Journal Name: Annals of Nuclear Energy Vol. 207; ISSN 0306-4549
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English

References (5)

Application of code scaling applicability and uncertainty methodology to the large break loss of coolant journal November 1998
AREVA's realistic large break LOCA analysis methodology journal July 2005
Full core LOCA safety analysis for a PWR containing high burnup fuel journal August 2021
CTF: A modernized, production-level, thermal hydraulic solver for the solution of industry-relevant challenge problems in pressurized water reactors journal October 2022
Rod Bundle Heat Transfer Thermal-Hydraulic Program journal July 2018

Figures / Tables (15)

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Table 1(p. 16)table Table 1
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Table 2(p. 27)table Table 2
Figure 11(p. 28)figure Figure 11
Figure 12(p. 29)figure Figure 12
Figure 13(p. 30)figure Figure 13

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Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FFRD
LOCA
Subchannel