Power-Cooling-Mismatch Test Series Test PCM-1: fuel rod behavior report
Description
The fuel rod behavior results of the Power-Cooling-Mismatch (PCM) Test Series, Test PCM-1 are presented. The objectives of the test were to evaluate the behavior of a pressurized water reactor (PWR) type fuel rod operated at high temperatures in film boiling for a time beyond rod failure with large local regions of molten fuel, and to evaluate the consequences of the continued operation. Test PCM-1, conducted in the Power Burst Facility, employed one 0.914-m-long PWR-design fuel rod, fabricated from unirradiated Zircaloy-4 and pressurized to 2.58 MPa with a fill gas of 77.7% helium and 22.3% argon. The rod was subjected to a preconditioning period and a power ramp to 69 kW/m rod peak power at a constant mass flux of 1110 kg/s.m2. The power was subsequently adjusted to 78 kW/m. The rod underwent film boiling during the power ramp to 69 kW/m and failed about 500 seconds later (about 470 seconds of film boiling operation at the peak temperature location). The test rod was in film boiling for 900 seconds with up to 85% of the local fuel pellet radius molten. The test information extends the existing PCM data base to include film boiling operation of a single fuel rod for a significant time beyond failure
Availability note (English)
MF available from INIS under the Report Number; Available from NTIS., PC E03/MF E03.
Files
Additional details
Additional titles
- Augmented title (English)
- PWR
Publishing Information
- Imprint Pagination
- 35 p.
- Report number
- NUREG/CR--0907
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 11517609
- Subject category
- S61: RADIATION PROTECTION AND DOSIMETRY; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- FILM BOILING; FUEL ELEMENT FAILURE; FUEL RODS; MELTDOWN; POWER-COOLING-MISMATCH ACCIDEN; PRESSURE GRADIENTS; PWR TYPE REACTORS; REACTOR SAFETY; SIMULATION; SPATIAL DISTRIBUTION
- Descriptors DEC
- ACCIDENTS; BOILING; DISTRIBUTION; FUEL ELEMENTS; PHASE TRANSFORMATIONS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SAFETY; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- TREE--1374.