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Simulation of Nuclear Fuel Rods by Electrically Heated Rods

Siegfried Malang, Klaus Rust

Nuclear Technology

Volume 58 / Number 1 / July 1982 / Pages 53-62


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For the investigation of thermohydraulic behavior during loss-of-coolant accidents (LOCAs), the nuclear fuel rods are simulated, in out-of-pile experiments, by electrically heated rods. These heater rods are required to produce temperature and heat flux histories at each position of the heater rod surface, identical to those of the nuclear fuel rods. Generally, these requirements are approximated by preprogramming of the transient heater rod power using estimated cooling conditions. However, the cooling conditions are not known very accurately prior to a test since the investigation of the thermohydraulics is the main purpose of the test. The use of an on-line process computer that controls the power of the heater rod by feedback of the measured cladding temperature to simulate, more closely, a LOCA has been suggested. A computer code simulating experiments in which the heater rod power is controlled by an on-line computer has been developed for checking and has demonstrated the validity of the method. In addition, the method has been confirmed by experiments performed at the Semiscale Test Facility.

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