Irradiation behavior of Zircaloy-clad fuel rods containing dished-end UO/sub 2/ pellets (LWBR Development Program)

Technical Report · Sat Jul 01 04:00:00 EDT 1967

OSTI ID:7010759

Berman, R M; Meieran, H B; Patterson, P W

Two fuel rods, BETT 79-21 and BETT 79-25, containing cylindrical UO/sub 2/ pellets with dished ends were irradiated in the B-4 Loop of the Materials Test Reactor to a depletion of 12.41 and 15.26 x 10/sup 20/ fissions/cc, respectively. Rod 79-21 was irradiated for nine days in the failed condition prior to removal; it showed extensive circumferential ridging of the clad at the interfaces between pellets, as well as hydriding, cracking, and spalling of the clad. Rod 79-25 was removed before failure when it also showed circumferential ridging of the clad. Approximately 35% of the total fuel swelling of a typical pellet was accommodated in the dishes; 49% more was accommodated in filling the internal porosity of the fuel material. The remaining 16% of the total swelling (combined with thermal expansion) is postulated to have caused the expansion of the diameter and ridging that led to failure. Total rate for all three types of fuel swelling was 0.82 v/o per 10/sup 20/ fission/cc at the conditions under which these samples were irradiated. (NSA 21: 39660)

• Research Organization: Bettis Atomic Power Lab., Pittsburgh, PA (USA)
• DOE Contract Number: AT(11-1)-GEN-14
• OSTI ID: 7010759
• Report Number(s): WAPD-TM-629
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500* -- Power Reactors
Breeding
ACTINIDE COMPOUNDS
ALLOYS
BREEDER REACTORS
CHALCOGENIDES
FUEL ELEMENTS
FUEL RODS
LWBR TYPE REACTORS
OXIDES
OXYGEN COMPOUNDS
PERFORMANCE
PHYSICAL RADIATION EFFECTS
RADIATION EFFECTS
REACTOR COMPONENTS
REACTORS
SWELLING
THERMAL REACTORS
TIN ALLOYS
URANIUM COMPOUNDS
URANIUM DIOXIDE
URANIUM OXIDES
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS