An Evaluation of the PBF LOFT Lead Rod Test Results Concerning Surface Thermocouple Perturbation Effects
The purpose of the Power Burst Facility Loss of Fluid Test (PBF LOFT) Lead Rod (LLR) Test program was to provide experimental data to characterize the mechanical behavior of LOFT type nuclear fuel rods under loss of coolant accident (LOCA) conditions, simulating the test conditions expected for the LOFT Power Ascension (L2) Test series. Although the LLR tests were not explicitly designed to evaluate cladding surface thermocouple perturbation effects, comparison of the Linear Variable Differential Transformer (LVDT) data for rods instrumented with and without cladding thermocouples provided pertinent information concerning the effects of cladding thermocouples on the time to DNB and time to quench data. Documentation and review of this data is presented in the following report. It will be shown that most of the LLR data indicate that the cladding surface thermocouples did not enhance the rewetting characteristics of the rods they are attached to, even though other evidence shows that the surface clad thermocouples did quench early. Finally, in order to accurately interpret and understand the limitations of the LVDT instrumentation, upon which thermocouple perturbation effects were evaluated, an analysis of the LVDT data as well as a review of the atypical response events that occurred during the LLR tests are presented in appendices to this document.
- Research Organization:
- Idaho National Engineering Laboratory (INEL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 1024876
- Report Number(s):
- L0-00-79-108; USNRC-P-394
- Country of Publication:
- United States
- Language:
- English
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