Modelling reactor core behaviour under accident conditions with delayed re-introduction of cooling water (commonly referred to as reflood) is a challenge for safety analysis computer codes. Reflood thermal-hydraulics (e.g. post-critical heat flux flow and heat transfer, entrainment, quench) remain a major contributor to code uncertainties in simulation of many accident scenarios and must be more deeply understood to enhance nuclear safety. Further, as the nuclear industry evolves, there is a need for additional data for power up-rates and new designs.
The objective of this three-year activity is to conduct new experiments and evaluate system hydraulics and sub-channel codes in the simulation of reflood tests in a full height rod bundle for complex inlet flows. While numerous experimental tests have been conducted for steady, constant inlet conditions, relatively few experiments have examined variable or oscillating inlet flows which are more likely in a hypothetical accident scenario.
To fill the knowledge gaps, tests will be performed in the Rod Bundle Heat Transfer (RBHT) facility at Pennsylvania State University under the US Nuclear Regulatory Commission (US NRC) co-ordination, with the objectives to:
In addition, test data from the project will be used to conduct a challenging benchmark exercise applying uncertainty methods as part of the code assessment process. The benchmark will comprise two phases, one with open tests data in 2020 and one with blind tests data in 2021.
The RBHT Project is supported by safety and research organisations in the following countries: Belgium, Czech Republic, Finland, France, Germany, Italy, Japan, Korea, Spain, Sweden, Switzerland and the United States.
October 2019 to October 2022
About EUR 1.44 million per year for the members
Last updated: 11 June 2020