• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Proceedings of the KWS Conference
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• 2007.11a
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• pp.203-205
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• 2007

A spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly for pressurized water reactors(PWRs). The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral loads acting on the nuclear fuel assembly so as to keep the nuclear fuel assembly straight. And also, the spacer grid assembly is hydraulically required to have less hydraulic resistance of coolant. To meet this requirement, it is necessary to weld the welding parts carefully and precisely. In this study, weld qualities such as, weld bead size and spatter manufactured by various welders were compared and analyzed. And performance parameters such as impact strength of spacer grid and hydraulic resistance of coolant were also compared and analyzed. Comparison results show that the weld qualities could be improved by selecting the optimal welding condition and also improving the welding technique.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.327-338
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• 2001

A computational fluid dynamics (CFD) analysis was performed to propose the optimum design of flow mixing vane on the space grid in a PWR fuel assembly. The flow mixing vanes considered in this study for optimum design are swirl-vane and twisted-vane. A single subchannel of one grid span was modeled using flow symmetry to minimize the computational effort. The CFD predictions are in good agreement with the experimental results for the split- vane, which shows the applicability of the CFD method. The mixing effect by swirling flow and crossflow, and the pressure drop were estimated and compared for the various vane angles. The optimum vane angle is proposed to be 40。 and 35。 from the direction of axial flow for the swirl-vane and the twisted-vane, respectively.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2396-2401
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• 2007

Freon CHF experiments are carried out to investigate the CHF enhancements by mixing vane shapes of spacer grids in nuclear fuel assembly. The experiments were performed for a wide range mass flux, 50$\sim}$3000 kg/$m^2s$. Three kinds of spacer grids in 5${\times}$5 rod bundles are tested: no mixing vane grids, hybrid mixing vane grids, and split mixing vane grids. The CHF performances are compared along with the data belong to the PWR operating conditions based on a water equivalence through a fluid-to-fluid modeling method. The average of the data in this range is 16.4% for 37 data of hybrid vane grid and 12.5% for 24 data of split vane. In the lower mass flux, however, the split vane grid shows slightly higher performance than the hybrid vane grid.

• Nuclear Engineering and Technology
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• v.39 no.6
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• pp.737-746
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• 2007

A spacer grid, which is one of the most important structural components in a PWR fuel assembly, supports its fuel rods laterally and vertically. Based on in-house design experience, scrutiny of the design features of advanced nuclear fuels and the patents of other spacer grids, KAERI has devised its own spacer grid shapes and acquired patents. In this study, a performance evaluation of KAERI's spacer grid shapes was carried out from mechanical/structural and thermohydraulic view points. A comparative performance evaluation of commercial spacer grid shapes was also carried out. The comparisons addressed the spring characteristics, fuel rod vibration characteristics, fretting wear resistance, impact strength characteristics, CHF enhancement, and the pressure drop level of the spacer grid shapes. The results show that the performances of KAERI's spacer grid shapes are as good as or better than those of the commercial spacer grid shapes.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.13-14
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• 2005

KAERI is performing a project on out-pile test technology development for a full scale PWR fuel assembly. We have introduced the hydraulic test facility, a test assembly, test parameters, test methods, and a data acquisition system. The start up test will be in the middle of March 2005 and the main test will be accomplished by the end of 2006. The established test facility and measuring technique will contribute to the satisfaction of domestic needs for the design verification to improve the reliability of a PWR plant operation.

• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.352-362
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• 1991

Recent popular trends in pressurized water reactor(PWR) fuel management are to extend the cycle length and to employ the low-leakage core designs for the optimal utilization of the uranium resources. In control strategy incorporated with the fuel management, turnable absorbers are required to control the power peaking and to ensure a negative moderator temperature coefficient during reactor operation. In this study, the nuclear characteristics and the optimal allocation of gadolinium-poisoned rods within the fuel assembly are considered using KWU SAV 79 A Code Package. First, analyses are carried out to compare the nuclear characteristics of the fuel assemblies contain-ing WABA(Wet Annular Burnable Absorber) and Gadolinium burnable absorbers respectively. The analyses show that the gadolinium-bearing fuel assembly has peculiar depletion characteristics ensuing from the very large thermal neutron absorption cross section. Peculiar characteristics of gadolinium provide basis for the optimal allocation of Gd rods in fuel assembly. Second, the methodology of an optimal allocation of gadolinium-poisoned rods within the fuel assembly is developed and applied to some nuclear designs.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.299-309
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• 1997

The neutronic feasibility of typical Korean three-loop 900MWe class PWR core loaded with mixed oxide fuels for both annual and 18-month cycle strategies has been investigated as a means for spent fuel management. For this study, a method of determining equivalent plutonium content was developed under the equivalence concept which gives the same cycle length as uranium fuel. Optimal plutonium zoning within the MOX assembly was also designed with the aim of minimizing the peak md power. Conceptual core designs hate hen developed for equilibrium cycle with the following variations: annual and 18-month cycle, 1/3 and full MOX loading schemes, and typical and high moderation lattice. The analysis of key core physics parameters shows that in all cases considered satisfactory core designs seem to be feasible, though addition of control rod system and change in Technical Specification for soluble boron concentration are required for full MOX loading in order to meet the current design requirements.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1854-1858
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• 2007

The reliability assurance with respect to the test procedure and results of the out-pile mechanical performance test for the nuclear fuel assembly is an essential task to assure the test quality and to get a permission for fuel loading into the commercial reactor core. For the case of vibration test, which is carried out to obtain basic dynamic characteristics of the fuel assembly, proper management and appropriate calibration of instruments and devices used in the test, various efforts to minimize the possible error during the test and signal acquisition process are needed. Additionally, the deep understanding both of the theoretical assumption and simplification cation for the signal processing/modal analysis and of the functions of the devices used in the test were highly required. Finally, to verify the test result to represent the accurate natural characteristics of the structure, the proper correlation analysis between the theoretical and experimental method has to be carried out. In this study, the overall procedure and result of lateral vibration test for the fuel assembly's mechanical characterization were briefly introduced. A series of measures to assure and improve the reliability of the vibration test were discussed.

• Proceedings of the KWS Conference
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• 2006.05a
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• pp.39-41
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• 2006

The weld quality of spacer grids in Pressurized Water Reactors(PWR) fuel is extremely important for the fuel assembly performance in the nuclear renter. The spacer grid welds are currently evaluated mainly by the metallographic examination although it reveals only cross-points which are welded by the laser beam. This experiment is also to compare the weldability of Zircaloy-4 spacer grids using by the GTA and laser beam. The effect of node geometries of spacer grids for GTAW and LBW has been studied and optimum conditions of spacer grid welding have been found. Microstructures and micro-hardness of GTA and laser beam welded zones have been also compared.