A study of transient FGR by integral LOCA tests

Conference · Fri Jul 01 04:00:00 EDT 2016

OSTI ID:22764080

Magnusson, P.; Jaedernaes, D. ^[1]; Adamsson, C.; Schrire, D. ^[2]; Roennberg, G. ^[3]; Alvestav, A. ^[4]; Seidl, M. ^[5]

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1. Fuel and Materials Technology, Studsvik Nuclear AB, Nykoeping, 611 82 (Sweden)
2. Vattenfall Nuclear Fuel, Solna 169 92 (Sweden)
3. Core and Fuel Engineering, OKG AB, Oskarshamn, 572 83 (Sweden)
4. Department of nuclear plant safety, Swedish radiation safety authority, Solna, 17116 (Sweden)
5. E.ON Kernkraft, Hannover, 30457 (Germany)

A study of transient fission gas release (FGR) was performed by means of two out-of-pile LOCA tests in a hot cell facility. Irradiated fuel rod segments with lengths of roughly 80 mm and burnups of 64 MWd/kgU were subjected to simulated water-cooled reactor LOCA transients in an integral LOCA test device. The segments were heated to a maximum temperature of 1000 Celsius degrees and the increase in rod inner pressure resulting from fission gas release and increased gas temperature during the transient was measured. The first release of gas occurred at temperatures close to 450 Celsius degrees, which suggests a burst release mechanism that may be linked to the onset of grain boundary fracturing. At the end of the simulated transients, gas samples were taken from the fuel rod segments for study by gas mass spectroscopy. The transient fission gas release was measured and amount to roughly 13 % of the calculated pre-test rod inventory of Kr and Xe. Isotopic ratios of the released gas indicated that transient FGR originated from the whole pellet and not only the pellet periphery. The tested segments were also studied by light-optical microscopy to determine pellet porosity, pellet crack pattern and fuel fragmentation. The extent of fine fuel fragmentation was found to be small and limited to the periphery of the fuel. The determined amount of transient fission gas release can be added to the database of fuel rod design codes to more accurately describe the LOCA cladding strains and burst temperatures in the future.

• Research Organization: American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22764080
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING
BURNUP
CLADDING
CRACKS
FISSION PRODUCT RELEASE
FRAGMENTATION
FUEL PELLETS
FUEL RODS
GRAIN BOUNDARIES
HOT CELLS
ISOTOPE RATIO
LOSS OF COOLANT
MASS SPECTROSCOPY
OPTICAL MICROSCOPY
POROSITY
REACTOR ACCIDENT SIMULATION
SPENT FUELS
STRAINS
WATER COOLED REACTORS