• Proceedings of the KSME Conference
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• 2003.11a
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• pp.1297-1302
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• 2003

KAERI has contrived 15 kinds of spacer grid shapes of its own since 1997 and applied for domestic and foreign patents. To date, KAERI has obtained US and ROK patents for 6 kinds of spacer grid shapes among them and the others are under review in USA, EC, China, and ROK. In this study, mechanical/structural performance analysis and test on two spacer grid shapes that are assumed to be the most effective candidates for the spacer grid of the next generation nuclear fuel in Korea was carried out. The result has shown that the performances of the candidates are better or not worse than those of the current spacer grid.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.432-437
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• 2004

KAERI has contrived 16 kinds of spacer grid shapes of its own since 1997 and applied for domestic and foreign patents. To date, KAERI has obtained US and ROK patents for 11 kinds of spacer grid shapes among them and the others are under review in USA, EC, China, and ROK. In this study, detailed performance analysis and test on two spacer grid shapes that are assumed to be the most effective candidates for the spacer grid of the next generation nuclear fuel in Korea was carried out. The result has shown that the performances of the candidates are better or not worse than those of the current spacer grid.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.318-329
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• 2016

Reducing critical boron concentration in a commercial pressurized water reactor core offers many advantages in view of safety and economics. This paper presents a preliminary investigation of a reduced-boron pressurized water reactor core to achieve a clearly negative moderator temperature coefficient at hot zero power using the newly-proposed "Burnable absorber-Integrated Guide Thimble" (BigT) absorbers. The reference core is based on a commercial OPR1000 equilibrium configuration. The reduced-boron ORP1000 configuration was determined by simply replacing commercial gadolinia-based burnable absorbers with the optimized BigT-loaded design. The equilibrium cores in this study were directly searched via repetitive Monte Carlo depletion calculations until convergence. The results demonstrate that, with the same fuel management scheme as in the reference core, application of the BigT absorbers can effectively reduce the critical boron concentration at the beginning of cycle by about 65 ppm. More crucially, the analyses indicate promising potential of the reduced-boron OPR1000 core with the BigT absorbers, as its moderator temperature coefficient at the beginning of cycle is clearly more negative and all other vital neutronic parameters are within practical safety limits. All simulations were completed using the Monte Carlo Serpent code with the ENDF/B-VII.0 library.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.87-93
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• 2006

Inventories, and characteristics such as dimension, fuel rod array, weight, $^{235}U$ enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial $^{235}U$ enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed $\sim36$ GWD/MTU and $\sim40$ GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be $\sim45$ GWD/MTU at the end of 2000's. From the comprehensive study, it was concluded that the imaginary SNF with $16\times16$ Korean Standard Fuel Assembly, cross section of $21.4cm\times21.4cm$, length of 453cm, mass of 672 kg, initial $^{235}U$ enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2431-2441
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• 2020

Currently, the pin-by-pin homogenized solvers are a very active research field as they can, unlike the nodal codes, directly predict the local power, while requiring significantly less computational resources than the heterogeneous transport codes. This paper presents a recently developed pin-by-pin diffusion/SP3 solver Tortin, its spatial discretization method and the reflector treatment. Regarding the spatial discretization, it was observed that the finite difference method applied on pin-cell size mesh does not properly capture the big flux change between MOX and uranium fuel, while the nodal expansion method is more accurate but too slow. If the finite difference method is used with a finer mesh in the outer two pin rows of the fuel assembly, it increases the required computation time by only 50%, but decreases the pin power errors below 1% with respect to lattice code reference solutions. The paper further describes the coupling of Tortin with a microscopic depletion solver. Several verification tests show that the SP3 pin-by-pin solver can reproduce the heterogeneous transport solvers results with very good accuracy, even for fuel cycle depletion of very heterogeneous core employing MOX fuel or inserted control rods, while being two orders of magnitude faster.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.337-344
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• 1994

Enthalpy rise hot channel factor($F_{\Delta{H}}$$^{N}$) is one of the most limiting constraints in determining the fuel loading pattern(LP) for PWR's. In order to enhance the LP design flexibility without any changes of not only basic fuel specifications but also Technical Specifications and Operation Procedures, we apply the enrichment zoning concept to Westinghouse designed PWR's to flatten the rod power distributions within the fuel assembly and thus to reduce $F_{\Delta{H}}$$^{N}$. Enrichment zoning is described that each assembly consists of two different enrichment fuels ; the lower enriched fuels are located in positions which are expected to have the higher rod power and vice versa for the higher enriched fuels. As a result of unit assembly calculations to flatten the rod power distribution within the assembly, the appropriate enrichment difference is found to be 0.3~0.4w/o. Through core depletion calculations for the 18-month cycle of Kori Unit 4, the $F_{\Delta{H}}$$^{N}$ behavior in core with the enrichment zoning concept is investigated. A comparison with the reference case without the enrichment zoning results in a reduction in $F_{\Delta{H}}$$^{N}$ of approximately 1.5%.TEX>H/$^{N}$ of approximately 1.5%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.225-232
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• 2008

An estimation has been made for inventories and characteristics of spent nuclear fuel(SNF) to be generated from existing and planned nuclear power plants based on the 3rd Basic Plan for Electric Power Demand and Supply. The characteristics under consideration in this study are dimensions, a fuel rod array, a weight, $^{235}U$ enrichment, and the discharge burnup in terms of fuel assembly. These are essentially needed for designing a pyroprocessing facility. It is appeared that the anticipated quantity by the end of 2077 is about 23,000 tU for PWR spent nuclear fuel. It is revealed that the proportion of SNF with the initial $^{235}U$ enrichment below 4.5 weight percent(wt.%) is approximately 95 % in total. For SNF with 16$\times$16 fuel rod array the proportion is expected approximately 74% in total. It appears that the average burnup of SNF will be 55 GWd/tU after the medium and/or latter part of 2010s while the average burnup is 45 GWd/tU at present. Finally, a requirement in terms of reference SNF for designing the pyroprocessing facility has been derived from the above-mentioned results. The anticipated SNF seems to be 16$\times$16 Korean Standard Fuel Assembly with a cross section of 21.4 cm$\times$21.4 cm, a length of 453 cm, a mass of 672 kg, the initial $^{235}U$ enrichment of 4.5 wt.%, and the discharge burnup of 55 GWd/tU.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.33-42
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• 2016

Spacer grids play an important role in maintaining the proper form of the fuel assembly structure and ensuring the safety of reactor core design. This study applies the Monte Carlo method to the analysis of the neutronics effects of spacer grids in a typical pressurized water reactor (PWR). The core problem used to analyze the neutronics effects of spacer grids is a modified version of Korea Advanced Institute of Science and Technology benchmark problem 1B, based on an Advanced Power Reactor 1400 (APR1400) core model. The spacer grids are modeled and added to this test problem in various ways. Then, by running MCNP5 for all cases of spacer grid modeling, some important numerical results, such as the effective multiplication factor, the spatial distributions of neutron flux, and its energy spectrum are obtained. The numerical results of each case of spacer grid modeling are analyzed and compared to assess which type has more advantages in accuracy of numerical results and effectiveness in terms of geometry building. The conclusion is that the most realistic modeling for Monte Carlo calculation is the "volume-preserving" streamlined heterogeneous spacer grids, but the "banded" dissolution spacer grids modeling is a more practical yet accurate model for routine (deterministic) analysis.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.483-490
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• 2012

PWR spent fuels produced in the Republic of Korea are expected to be recycled by pyroprocess in the long term future. Even though pyroprocess waste amounts can be smaller than that of PWR spent fuel assembly in case of direct disposal, this process essentially will produce various and unique radioactive wastes. The goals of this article are to characterize these wastes, calculate the amount of wastes, design disposal systems for each waste and evaluate the radiation safety of each system by dose assessment. The absorbed dose results of the metal and ceramic waste for the engineering barrier system (EBS) showed $2.21{\times}10^{-2}$ Gy/h and $1.15{\times}10^{-2}$ Gy/h, which are lower than the recommended value of 1 Gy/h. These results confirmed that the newly proposed disposal systems have a safety margin for the radiation produced from each waste.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2002.05a
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• pp.135-141
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• 2002

An optimization of KTF thorium fuel assembly design was performed on the basis of the design parameter studies. Optimization goals ware to make the core have both proliferation resistance and fuel cycle economics. Four kinds of proliferation resistance indexes were used; SNS, TG, BCM, Toxicity. A new index, FEI was regarded as a limiting index for the maximization of fuel cycle economics. Optimized thorium fuel design was applied for APR-1400 reactor core. Nuclear core design procedures were examined to solve the thorium fuel reactor problems. It was shown that heterogeneous thorium fuel core option is acceptable in safety and economics aspects.