Irradiation performance of HTGR fuel rods in HFIR experiment HRB-6

Technical Report · Mon Dec 01 04:00:00 EST 1975

DOI:https://doi.org/10.2172/4134624· OSTI ID:4134624

Homan, F. J.; Long, Jr., E. L.; Montgomery, B. H.; Hamner, R. L.; Valentine, K. H.

The HRB-6 Capsule was irradiated in the RB-5 facility of the High Flux Isotope Reactor (HFIR) for eight cycles. The primary purpose of this experiment was to test the then reference recycle fissile fuel--Bisco coated mixed thorium- uranium oxide. Among the several secondary objectives was to determine whether there was any difference in irradiation performance of $sup 233$U and $sup 235$U. Experiment HRB-6 was the first test of fully enriched (93 percent) $sup 235$U in HFIR tests. Both slug injection and extrusion fabrication techniques were used to prepare specimens for this test. No detectable difference between $sup 233$U and $sup 235$U irradiation performance could be seen from this experiment. The migration rates for the mixed oxide kernels measured for the HRB-6 specimens were consistent with such measurements made on similar kernels in other experiments. Analysis of the entire body of data on thermal migration for mixed oxide fissile kernels later led to the conclusion that this fuel was a marginal performer, and ultimately was replaced as the reference fuel by a uranium oxycarbide fissile kernel loaded from ion exchange resins. The observation of identical performance for $sup 233$U and $sup 235$U in the mixed oxide system was very important to the fuel development program. The recycle fuel development efforts have continued to use $sup 235$U in subsequent irradiation tests, at considerably less expense than if $sup 233$U test specimens were used. The very good performance of the test specimens fabricated using extrusion, relative to those fabricated using slug injection, was noted again in HRB-6. (auth)

• Research Organization: Oak Ridge National Lab., Tenn. (USA)
• DOE Contract Number: W-7405-ENG-26
• NSA Number: NSA-33-008700
• OSTI ID: 4134624
• Report Number(s): ORNL-TM--5011
• Country of Publication: United States
• Language: English

Related Subjects

*FUEL ELEMENTS-- TESTING
*HTGR TYPE REACTORS-- FUEL ELEMENTS
210300* --Nuclear Power Plants--Power Reactors
Non- Breeding
Graphite Moderated
FISSION PRODUCTS
HFIR REACTOR
N77300* --Reactors--Power Reactors
Non-breeding
Graphite Moderated
PHYSICAL RADIATION EFFECTS
THORIUM OXIDES
URANIUM 233
URANIUM 235
URANIUM OXIDES