Laboratory data for review of outlet water temperature limits for BDF type reactors
A knowledge of the thermal and hydraulic conditions within a reactor fuel channel during an inadvertent flow reduction is needed to establish reactor operating limits. Such limits, which are based on outlet water temperature, are called ``trip-after-instability`` limits by the reactor operating personnel. Laboratory experiments were performed to update the knowledge of such conditions in a BDF reactor type fuel channel while using internally and externally cooled fuel elements (I&E`s) at tube powers up to 1530 KW. In addition to a general extension of previous data, the new data were used to review certain specific details involved in ``trip-after-instability`` limit calculations. It was found that in calculating the limits, the isothermal pressure drop across the fuel elements must be related to flow rate by the exponent 1.8, ({delta}P {proportional_to} F{sup 1.8}), rather than the more convenient value of 2.0. It was found that this method of limit determination is applicable to the high rear header pressures presently attained on the reactors and also applicable to tubes with very low Panellit pressures. And finally, the validity of certain analytical transformations of experimental data, called generalization of hydraulic demand curves, was reaffirmed for the above conditions.
- Research Organization:
- General Electric Co., Richland, WA (United States). Hanford Atomic Products Operation
- Sponsoring Organization:
- USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC06-76RL01830
- OSTI ID:
- 10122330
- Report Number(s):
- HW-67139; ON: DE93006715
- Resource Relation:
- Other Information: PBD: 11 Dec 1960
- Country of Publication:
- United States
- Language:
- English
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