Jae-Yong Oh

DROP IMPACT ANALYSIS OF PLATE-TYPE FUEL ASSEMBLY IN RESEARCH REACTOR

In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to determine whether the fuel plate integrity is maintained in a drop accident. A fuel assembly drop accident is classified based on where the accident occurs, i.e., inside or outside the reactor, since each occasion results in a different impact load on the fuel assembly. An analysis procedure suitable for each drop situation is systematically established. For an accident occurring outside the reactor, the direct impact of a fuel assembly on the pool bottom is analyzed using implicit and explicit approaches. The effects of the key parameters, such as the impact velocity and structural damping ratios, are also studied. For an accident occurring inside the reactor, the falling fuel assembly may first hit the fixing bar at the upper part of the standing fuel assembly. To confirm the fuel plate integrity, a fracture of the fixing bar should be investigated, since the fixing bar plays a role in protecting the fuel plate from the external impact force. Through such an analysis, the suitability of an impact analysis procedure associated with the drop situation in the research reactor is shown.

FUEL PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO₂ AND MOX FUEL

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in UO2 fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS’s precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code’s prediction. The database consists of the UO2 irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and UO2 fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

SIMULATION OF HIGH BURNUP STRUCTURE IN UO₂ USING POTTS MODEL

The evolution of a high burnup structure (HBS) in a light water reactor (LWR) UO₂ fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the UO₂ matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels.

CONSERVATIVE WIDTH OF HIGH-BURNUP STRUCTURE IN LIGHT WATER REACTOR UO 2 FUEL AS A FUNCTION OF PELLET AVERAGE BURNUP

Abstract Based on the high-burnup fuel data available in open literature, a conservative width of high-burnup structure (HBS) in light water reactor UO2 fuel, which can be used for fuel performance and accident analysis or assessment of spent fuel under geological disposal conditions, is proposed as a function of pellet average burnup. For pellet average burnup of 30 to 60 GWd/t U, where the HBS generally increases with burnup because of the accumulation of irradiation damage, a conservative HBS width is given by wHBS = 13.3 (buavg - 30), where wHBS is the HBS width in μm and buavg is the pellet average burnup in GWd/t U. For pellet average burnup of 60 to 75 GWd/t U, where microstructural damage caused by irradiation is partly annealed, a conservative HBS width is expressed by wHBS = 2.02 exp(buavg/11.35). In the case of pellet average burnup above 75 GWd/t U up to at least 100 GWd/t U, the HBS width does not exceed some limiting value of 1.5 mm because high temperature in the central region of the fuel pellet has caused an extensive annealing of irradiation damage. In addition, because of significant fission gas release during irradiation up to high burnup, HBS formation might not have expanded to the pellet region whose temperature was lower than the threshold one. Therefore, for this burnup range, a conservative HBS width is given as wHBS = 1500 μm.

Irradiation Test of MOX Fuel Rods Fabricated by Attrition-Milling and Analysis of In-Pile Data with COSMOS Code

Abstract Attrition-milling technology for fabricating mixed oxide (MOX) fuel was developed to mix the plutonium in UO2 fuels as homogeneously as possible. The fabricated MOX fuels were instrumented with temperature and pressure gauges that enabled one to measure the fuel temperature and rod internal pressure online. An irradiation test in the Halden reactor was performed to investigate the in-pile behavior of the fabricated MOX fuel. The irradiation of 1020 effective full-power days was successfully accomplished with good integrity of the test fuel rods. The rod average burnup reached ~50 MWd/kg HM, and the measured fuel centerline temperature was ~1000°C for the MOX fuels. A significant fission gas release was observed due to the high power level. The online measured in-pile performance data of the two attrition-milled MOX fuel rods were analyzed and compared with the fuel performance code COSMOS. COSMOS simulated the fuel centerline temperature and rod internal pressure for both MOX fuel rods. The analysis by COSMOS showed good agreement with the online measured in-pile behavior of MOX fuel.