UO$sub 2$--PuO$sub 2$ fuel rod bundle irradiation Mol-7A evaluation of the investigation results on the irradiated fuels rods
Work performed under United States -- Euratom Fast Reactor Exchange Program. In the irradiation test Mol-7A, seven UO/sub 2/-20% PuO/sub 2/ fuel rods (500 mm fuel length) were irradiated in a fuel rod bundle cooled by flowing sodium. Irradiation was performed in the epithermal neutron flux of the test reactor BR 2 (Mol) at rod powers between about 300 W/cm (fuel end) and 560 W/cm (neutron flux maximum) to a maximum burnup of about 45 MWdMg (U + Pu). The maximum fuel center temperature was calculated to be about 2800 deg C in the beginning, and about 2000 deg C for the largest part of irradiation time. The maximum cladding i.d. temperature (at the hot rod ends) was between 570 and 620 deg C. The amount of fission gas was measured in three steps as released fission gas, fission gas retained in large bubbles and fission gas retained in small bubbles and crystal lattice sites. The void volume distribution in fuel rod cross-sections was investigated by measuring gap widths, crack widths, central hole diameters, and porosity. The measured results were compared to computations by means of the fuel rod performance modeling-code SATURN. Electron microprobe analysis was applied to examine plutonium redistribution, fissionproduct compound composition and distribution and cladding corrosion. Redistribution of urarium and plutonium in the radial fuel temperature gradient results in Pu-enrichment around the central void amounting up to 28% PuO/sub 2/ (original content 20%) at the top (hot) end of the fuel column. The radial distribution of the fissionproduct Nd was measured in fuel solid solution. A large number of fission product metal and oxide phases could be found, which partly contained also Fe -- Ni (metal phases) and Cr (oxide phases) in the gap and on the fuel surface. The radial distribution of metallic Mo --Tc --Ru --Rh- Pd precipitates exhibited maximum precipitation density in the cooler part of the columnar garin zone. Tellurides and Pd-containing phases with volatile fission products (Sn, Sb, Te) exist preferably in the porous outer fuel zone and in the gap. Ba --molybdate, - zirconate, and -cerate were occasionally observed in the fuel and in the gap. Cs- oxide, -molybdate, -chromate, -uranate and -plutonate had been formed only in the cooler part of the gap. Chemical reaction between the UO/sub 2/ -- PuO/sub 2/ fuel (original O/M = 1.98 plus or minus 0.015) and the cladding tubes of three different austenitic stainless steels was shown by micrographs and microprobe measurements. The reaction became observable at a cladding i.d. temperature of about 500 deg C. The changes of cladding mechanical properties under irradiation were measured by stressrupture tube burst tests and by tensile tests on fuel rod sections. The tensile test results showed usual irradiation strengthening. The stress-rupture test results allow the conclusion that the cladding tubes were weakened over their full length. No dependence on the fast neutron fluence (0.5 - - 7 x 10/sup 21/ n/cm/sup 2/) could be found, but a clear decrease of the remaining rupture-time with increasing cladding irradiation temperature. The cladding material ss 1.4988 retained higher stress-rupture burst strength than 1.4961 and AISI 316. (auth)
- Research Organization:
- Kernforschungszentrum Karlsruhe (F.R. Germany). Inst. fuer Material- und Festkoerperforschung
- NSA Number:
- NSA-29-023500
- OSTI ID:
- 4340700
- Report Number(s):
- EURFNR-1113
- Resource Relation:
- Other Information: Work performed under United States--Euratom Fast Reactor Exchange Program. Orig. Receipt Date: 30-JUN-74
- Country of Publication:
- Germany
- Language:
- English
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*FUEL RODS- TESTING
*LMFBR TYPE REACTORS- FUEL RODS
*PLUTONIUM OXIDES- PHYSICAL RADIATION EFFECTS
*STAINLESS STEELS- PHYSICAL RADIATION EFFECTS
*URANIUM DIOXIDE- PHYSICAL RADIATION EFFECTS
FUEL ELEMENT CLUSTERS
FUEL-CLADDING INTERACTIONS
IRRADIATION
TEMPERATURE DEPENDENCE