• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.70-75
• /
• 1997

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14$\times$14 PWR fuel assembly and the results are given in this paper.

• Nuclear Engineering and Technology
• /
• v.55 no.7
• /
• pp.2591-2603
• /
• 2023

Simulating the Corrosion-Related Unidentified Deposit (CRUD) on the surface of fuel assemblies is necessary to predict the axial offset anomaly and the localized corrosion induced by the CRUD during the operation of nuclear power plants. A new CRUD model was developed to predict the formation of the CRUD deposits, considering the deposition and erosion mechanisms. The heat transfer and capillary flow within the CRUD were also considered to evaluate the boiling amount within the CRUD layer. This model predicted a CRUD deposit thickness of 44 ㎛ during a one-cycle operation of the Seabrook nuclear power plant. The CRUD deposition tended to accelerate and decelerate during the simulation, by being related to boiling mechanism on the deposits surface. Additionally, during a three-cycle operation corresponding to the refueling period, the CRUD deposition was saturated at a thickness of 80 ㎛, which was in good agreement with the suggested thickness for CRUD buildupin pressurized water reactors. Surface boiling on the thin CRUD deposits enhanced the acceleration of the deposition, even when the wick boiling properties were not favorable for CRUD deposition. To ensure the certainty of the simulation results, sensitivity analyses were conducted for the porosity, chimney density, and the constants employed in the proposed model of the CRUD.

• Nuclear Engineering and Technology
• /
• v.25 no.2
• /
• pp.248-254
• /
• 1993

The purpose of this study is to evaluate the thermal safety for dry transport of a shipping cask. Analysis condition was based on an ambient temperature of 38$^{\circ}C$ for normal heat condition. The cask was designed to carry 4PWR spent fuel assemblies with a burnup of 38,000 MWD/MTU and 3 years of cooling time. Thermal analysis was carried out by using the COBRA-SFS code. The fuel cavity was considered to be filled with air, nitrogen or helium gas for dry transport. The results of analysis showed that the maximum temperatures of fuel rod cladding in air and helium cavity would be 277$^{\circ}C$ and 226$^{\circ}C$, respectively, for 3 years of cooling time. These values were less than the specified temperature to maintain the thermal integrity of fuel assembly for dry transport.

• Nuclear Engineering and Technology
• /
• v.20 no.3
• /
• pp.145-154
• /
• 1988

This paper is a study on an accurate and computationally efficient method for reconstructing pointwise power distributions from coarse-mesh nodal calculations. The modern nodal codes can calculate global reactor power shapes and criticality very efficiently and accurately. But inherent in the nodal procedures, there is inevitable loss of information on local heterogeneous quantities. In this study, an improved form function method which reflects the exponential transition of the thermal flux near the assembly surface is developed for the reconstruction of the heterogeneous fluxes. Use of the new form function method in several pressurized water reactor (PWR) benchmark problems reduces the maximum errors in the reconstructed thermal flux to those in the reconstructed fast flux. Even for assemblies adjacent to the steel baffle in realistic PWR cores, use of this method also results in improved pointwise power reconstruction.

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.1907-1916
• /
• 2020

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05a
• /
• pp.21-26
• /
• 1997

Recently, thorium is receiving increasing attention as an important fertile material for the expanding nuclear power programs around the world. The superior nuclear and physical properties of thorium-based fuels could lead to very low fuel cycle cost and make thorium reactors economically attractive. In addition, the use of thorium in reactors would permit more efficient utilization of low cost uranium reserves and reduction nuclear wastes. In this work, the nuclear characteristics of a new type thorium fueled reactor (Radkowsky Thorium Reactor) consisting seed-blanket assemblies are addressed and compared with those typical assemblies of a PWR (CE type). Also, an assessment on several advantages of thorium fueled reactors is provided. All these results are based on the HELIOS code calculation.

• Nuclear Engineering and Technology
• /
• v.26 no.2
• /
• pp.197-204
• /
• 1994

In a PWR rod cluster control assembly(RCCA) for shutdown is released upon action of control rod drive mechanism and falls down through the guide thimble by its weight. Drop time and impact velocity of the RCCA are two key parameters with respect to reactivity insertion time and the mechanical integrity of fuel assembly. Therefore, the precise control of drop time and impact velocity is prerequisite to modifying the existing design features of the RCCA and guide thimble or newly designing them. During its falling down into the core, the RCCA is retarded by various forces acting on it such as fluid resistance caused by the RCCA movement, buoyance and mechanical friction caused by contacting inner surface of the guide thimble, etc. However, complicated coupling of the various forces makes it difficult to derive an analytical dynamic equation for the drop time and impact velocity. This paper deals with the development of a computer program containing an analytical dynamic equation applicable to the Korean Fuel Assembly(KOFA). The computer program is benchmarked with an available single control rod drop tests. Since the predicted values are in good agreement with the test results, the computer program developed in this paper can be employed to modify the exiting design features of the RCCA and guide thimble and to develope their new design features for advanced nuclear reactors.

• Proceedings of the KSME Conference
• /
• 2001.06c
• /
• pp.548-553
• /
• 2001

Recently, much attention is imposed on the design of the fuel assemblies in the Pressurized Light Water Reactor (PWR). Spacer grid is one of the main structural components in a fuel assembly. It supports fuel rods, guides cooling water and protects the system from the external impact loads. Various space grids have been proposed and new designs are also being created. In this research, a new spacer grid is designed by the axiomatic approach. The Independence Axiom is utilized for the design. For conceptual design, functional requirements (FRs) are defined and corresponding design parameters (DPs) are found to satisfy FRs in sequence. Overall configuration and shapes are determined in this process. Detail design is carried out based on the result of the axiomatic design. For the detail design, the system performances are evaluated by using linear and nonlinear finite element analysis. The dimensions are determined by optimization. Some commercial codes are utilized for the analysis and design.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.26 no.8
• /
• pp.1623-1630
• /
• 2002

Recently, much attention is imposed on the design of the fuel assemblies in the Pressurized Light Water Reactor (PWR). Spacer grid is one of the main structural components in a fuel assembly. It supports fuel rods, guides cooling water, and maintains a coolable geometry from the external impact loads. In this research, a new shape of the spacer grid is designed by the axiomatic approach. The Independence axiom is utilized for the design. For conceptual design, functional requirements (FRs) are defined and corresponding design parameters (DPs) are found to satisfy FRs in sequence. Overall configuration and shapes are determined in this process. Detail design is carried out based on the result of the axiomatic design. For the detail design, the system performances are evaluated by using linear and nonlinear finite element analysis. The dimensions are determined by optimization. Some commercial codes are utilized for the analysis and design.

• Proceedings of the KSME Conference
• /
• 2003.04a
• /
• pp.431-437
• /
• 2003

KAERI has contrived 14 kinds of spacer grid shapes of its own since 1997 and applied for Korean and US patents. To date. KAERI has obtained US and Korean patents for 6 kinds of spacer grid shapes among them. Tn this study. performance 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 Qui through the mechanical/structural test and analysis. The test result has shown thai the performances of the candidates are better or not worse than that of the current spacer grid.