Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors

  • Hossam H. Abdellatif (University of Idaho) ;
  • Palash K. Bhowmik (Idaho National Laboratory) ;
  • David Arcilesi (University of Idaho) ;
  • Piyush Sabharwall (Idaho National Laboratory)
  • Received : 2023.08.31
  • Accepted : 2024.01.30
  • Published : 2024.06.25
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Abstract

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

Keywords

Acknowledgement

This research was funded by United State (U.S.) Department of Energy (DOE) Advanced Reactor Demonstration Project (ARDP) program office grant number ARDP-20-23819. Funding Opportunity Number DE-FOA-0002271, Risk Reduction Pathway. The authors would like to thank the U.S. DOE National Reactor Innovation Center (NRIC), ARDP program office, and Irradiation Experiment and Thermal Hydraulics Analysis Department at Idaho National Laboratory (INL) for the encouragement and support.

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