Transactions of the Korean Society of Pressure Vessels and Piping (한국압력기기공학회 논문집)

Korean Society of Pressure Vessels and Piping (한국압력기기공학회)
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Drop Performance Test of Control Rod Assembly for Prototype Gen-IV Sodium-cooled Fast Reactor

PGSFR 제어봉집합체 낙하성능시험

  • Received : 2016.06.01
  • Accepted : 2016.06.17
  • Published : 2016.06.30
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Abstract

The Control Rod Assembly (CRA) controls the reactor power by adjusting its position in the reactor core during normal operation and should be quickly inserted into the reactor core by free drop under scram condition to shut down chain reactions. Therefore, the drop time of the CRA is one of important factors for the safety of the nuclear reactor and must be experimentally verified. This study presents the drop performance test of the CRA which has been conceptually designed for the Proto-type Generation IV Sodium-cooled Fast Reactor. During the test, the CRA was free dropped from a height of 1 m under different flow rate conditions and its drop time was measured. The results showed that the drop time of the CRA increased as the flow rate increased; the average drop times of the CRA were approximately 1.527 seconds, 1.599 seconds and 1.676 seconds at 0%, 100% and 200% of design flow rates, respectively.

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References

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