Results of Test lE-5 in the Irradiation Effects Test Series• administered under the Thermal Fuels Behavior Program of EG&G Idaho, Inc. for the Nuclear Reguia.t@ey Commission are presented. the objectives of this test were to evaluate the influence of simulated fission products, cladding irradiation d~age, and fuel rod internal pressure on pellet-c_ladding interaction during a power ramp ana on fuel rod behavior during film boiling operation. Test IE-5, conducted in the Pow~r Burst Facility at•the Idaho National Engii1eedrtg Laboratory, employed three 0.97-m long pressurized water reactor type fuel rods, fabricateci from previously irradiated zircaioy-4 claddmg and one similar rod fabricated ti:'om unirradiated cladding. The four rods were subjected to a preconditioning period, •a •power ramp to an .average fuel rod .peak power of 65 kW /m, and steady state.operation for one hour at a coolant mass flux of 4880 kg/s~m 2 for each rod. After a flow reduction to 1800 kg/s-m 2 , film boiling occurred on one rod. Additional flow reductions to 970 kgfs..:m? produced film boiling on the three remaining fuel rods. Maximum time in film boiliflg •was 80 s. The rod having the highest initial internal pressure (8.3 MPa) failed 10 s after the ons€t of film boiling. A second rod failed about 90 s after reactor shutdown. This report contains a description of the experiment; the test conduct, test results, .and results from the preliminary postirradiation examination. Calc]Jlations using a transieilt fuel rod behavior code ate compared with the test results.
The accident at Three Mile Island Unit-2 (TMI-2) on March 28, 1979 caused extensive damage to the core. A variety of analyses were performed using three general approaches to determine the extent of core damage. First, thermal-hydraulic events were reconstructed using available data, thermal-hydraulic principles, and computer analyses. Second, determinations of tne hydrogen generated yielded estimates of tne amount of zircaloy oxidized and embrittled. Third, the type and quantity of fission products released during the accident were used to estimate the location of core damage and the fuel temperatures which were achieved. Uncertainties exist in each type of determination due to the equivocal nature of the data. Thus, the purpose of this paper is to review and summarize the core damage assessments which have been made, identify the minimum and maximum bounds of damage, and establish a "reference" description for the current status of the core. The different degrees of damage present in the reference core will be considered during development of contingency tooling and procedures for inspection, sample acquisition, and defueling of the core. From reconstruction of the thermal-hydraulic events, it was concluded that the core remained covered up to 100 minutes into the accident and that most of the damage occurred during the period from 100 to 210 minutes when the core is thought to have been uncovered. Damage to the core is a strong function of the time-dependent steam-water mixture level in the core which was greatly affected by the net makeup flow during this period. Cladding reached temperatures between 1030 K (1395°F) and 1150 K (1610°F) and failed by ballooning between 137 and 142 minutes. The cladding continued to increase in temperature, becoming oxidized and embrittled. The fuel reached peak temperatures varying between 2000 K (3140°F) and 2900 K (4760°F). At temperatures greater than 2175 K (3455°F), uranium dioxide fuel pellets in contact with molten cladding could nave been dissolved by the zircaloy, forming a liquid phase of zirconium-uranium-oxygen termed "liquified fuel." The temperatures achieved are also high enough to melt 1 \/
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This report describes the results of Test IE-3, the fifth test in the Irradiation Effects Test Series conducted under the Thermal Fuels Behavior Program of EG&G Idaho, Inc. for the Nuclear Regulatory Commission. The objectives of this test were to: (a) determine the behavior of irradiated fuel rods subjected to a •rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four, 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. . The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m 2. After a flow reduction to 2120 kg/s-m 2 , film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as• a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions.
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