INVESTIGATION OF PLUTONIUM-BEARING FUEL ELEMENTS FOR THE PLUTONIUM RECYCLE TEST REACTOR. PART I

Technical Report · Thu Jun 01 00:00:00 EDT 1961

OSTI ID:4772001

Bailey, W J

Ten of the irradiation experiments with Pu-containing fuel elements which have been or are being conducted in support of the PRTR are described. Fuel elements containing Al--Pu and Al-12 Si--Pu alloys with 1.65 to 20 wt % Pu and high- and low-density, UO/sub 2/--PuO/sub 2/ pellets with 0.0259 to 7.45 at.% PuO/sub 2/ were fabricated and irradiated. Eight of the irradiation tests were performed in the Materials Testing Reactor (MTR) with capsules containing fuel cores which were 0.5 in. in dia. and 2.0 in. in length. Two tests were carried out in the high-temperature and pressure loops at the KER Facility at HAPO. One test employed a four-rod cluster element about ten in. long and one a seven-rod cluster which was 28 in. long. In both clusters the fuel cores were 0.5 in. in dia. The fuel element cladding was either Zircaloy-2 or Zircaloy-3 and had an outside dia. of 9/16 in. and a wall thickness of 0.030 in. All fuel elements employed in the tests performed satisfactorily. One irradiated high-density UO/ sub 2/--PuO/sub 2/ capsule failed when it was inadvertently placed in a neutron flux 11 times the specified value. Except for the ruptured specimen, all pieces were in good condition and exhibited no excessive or unusual corrosion. Core- core bonding was noted on one Al--Pu alloy fuel rod and core-end cap bonding on one Al--Si-- Pu alloy rod. From the limited number of experiments and observations it was concluded that the Al-- Pu alloys, Al-- Si-- Pu alloys, and U0/sub 2/-- PuO/sub 2/ compositions tested all exhibited good radiation behavior in the ranges of interest to the PRTR. For the same temperature range (650 to 850 deg F), the Al--Pu alloys appear to be more resistant to deformation than the Al--Si-- Pu alloys. In both alloy cases, resistance to deformation tends to increase with increasing Pu content. The volume changes appear to be more sensitive to core temperature than to exposure. The cluster experiments indicate that the PRTR element design is sound and will perform well in highpressure and temperature water. The oxide specimens received exposures on the order of 1000 to 2000 Mwd/t. Grain growth was noted in one high-density piece and a central void formation was observed in two, low-density pieces. The fission gas release was up to 3% in the highdensity oxides and to about 22% in the low-density specimens. The degree of fuel core cracking appears to decrease with increasing Pu content. Capsules from these tests are currently being irradiated to higher exposures (5000 Mwd/t). The oxide capsule which failed operated in the high flux position for six days before it was removed. All of the oxide was missing and a marked core-end cap reaction had occurred. The examination of the thin-wall, extruded Zircaloy tubing indicated that all material must be carefully inspected for flaws, especially on the inner surface. It is recommended that an ultrasonic technique be employed. (auth)

• Research Organization: General Electric Co. Hanford Atomic Products Operation, Richland, Wash.
• DOE Contract Number: AT(45-1)-1350
• NSA Number: NSA-16-030075
• OSTI ID: 4772001
• Report Number(s): HW-70158(Pt.I)
• Country of Publication: United States
• Language: English

Related Subjects

ALUMINUM ALLOYS
BONDING
CAPSULES
CONFIGURATION
CORROSION
DEFORMATION
DENSITY
EXTRUSION
FAILURES
FISSION PRODUCTS
FUEL CANS
FUEL ELEMENTS
FUELS
GASES
HIGH TEMPERATURE
IN PILE LOOPS
INSPECTION
IRRADIATION
LEAKS
MATERIALS TESTING
MIXING
PELLETS
PERFORMANCE
PLANNING
PLUTONIUM
PLUTONIUM ALLOYS
PLUTONIUM OXIDES
POWER PLANTS
PRESSURE
PRTR
QUANTITY RATIO
REACTOR TECHNOLOGY
REACTORS
SILICIDES
SURFACES
THICKNESS
ULTRASONICS
URANIUM DIOXIDE
VARIATIONS
VOLUME
ZIRCALOY
ZONES