• Journal of the Korean Society of Safety
• /
• v.30 no.3
• /
• pp.107-113
• /
• 2015

The goal of this paper is to investigate structural integrity factors of RMI(reflective metal insulation) to confirm the design requirements in nuclear power plant. Currently, a glass wool insulation is using now, but it will gradually be replaced with the reflective metal insulation maded by stainless steel plates. The main function of an insulation is to minimize a heat loss of vessel and pipes in RCS(reactor coolant system). It has to maintain structural a integrity in nuclear power plant life duration. In this study, the structural integrity analysis was carried out both multi-plate and outer shell plate by using a static analysis and experimental test. First, inner multi-plate has a self support structure for being air space. Because the effect of total static weight in multi-layer plate is low, a plate collapse possibility is not high. Considering optimum thin plate pressing process, it has to pre-check the basic physical properties. Second, the outer segment thickness and stiffener shape are verified by the numerical static analysis, and sample test for both type of panel and cylindrical pipe model.

• Nuclear Engineering and Technology
• /
• v.27 no.3
• /
• pp.327-335
• /
• 1995

The normal shutdown and maintenance period of a nuclear power plant can be remarkably shortened when the examination and maintenance works in steam generator tubes are simultaneously carried out with refueling job. There are nozzle dams to Hock the coolant How from reactor to steam generator. Workers are reluctant to install nozzle dam because of the high radiation exposure and the limited working space in steam generator. Moreover, the heavy weight of present nozzle dam makes it installation and removal works much difficult. In this paper, a lighter KAERI nozzle dam with increased flexural rigidity-to-weight was designed and manufactured by changing the structure design of the present nozzle dam and by selecting new material, carbon fiber-reinforced plastic.

• Nuclear Engineering and Technology
• /
• v.24 no.3
• /
• pp.228-235
• /
• 1992

Some soluble species may not be solubility-limited or congruent-released with the matrix species. For example, during the operation of the nuclear reactor, the fission products can be accumulated in the fuel-cladding gap, voids, and grain boundaries of the fuel rods. In the waste package for spent-fuel placed in a geologic repository, the high solubility species of these fission products accumulated in the“gap”, e.g. cesium or iodine are expected to dissolve rapidly when ground water penetrates fuel rods. The time and space dependent mass transport for high solubility nuclides in the gap is analyzed, and its numerical illustrations are demonstrated. The approximate solution that is valid for all times is developed, and validated by comparison with an asymptotic solution and the solution obtained by the numerical inversion of Laplace transform covering the entire time span.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.20 no.3
• /
• pp.215-222
• /
• 2010

Long and slender body, like a fuel rod, oscillating in axial flow can be unstabilized even by the small cross flow which can be activated by the flow mixer or turbulent generator. It is important to include these effects of flow disturbance in dynamic stability analysis of nuclear fuel rod. This work shows how eigen frequency of a multi-span fuel rod can be changed by the swirl flow, which is discretely generated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was calculated. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
• /
• v.19 no.3
• /
• pp.192-197
• /
• 1987

Two-point kinetic equations for a compact-core-with-bulky-D$_2$O-reflector system were developed. A unique feature of the system is that certain fission gammas create retarded photoneutrons in the D$_2$O reflector by (ｒ, n) reaction. Coupling effect between the core and the reflector was investigated by simulating power transients with various ramp reactivity insertions. Special attention was paid to the phenomenon associated with spatial separation of photoneutrons and their precursors. Simulations show that accuracy of the two-point model is comparable with that of space-dependent approach. Also it is found that the explicily expressed photoneutron terms in the reflector equation slow down the power transient compared to non-photoneutron expressions. Detectors for reactor power control purpose prefer to be deployed in the core zone to be able to accurately perdict transient power.

• Nuclear Engineering and Technology
• /
• v.23 no.2
• /
• pp.183-199
• /
• 1991

Transient response of temperature distributions in a Pressurized Water Reactor (PWR) pressurizer vessel wall for the Inadvertent Auxiliary Spray transient has been analyzed with conservatism accounted for the resulting thermal stresses in the regions of the vessel wall which are wetted by the spray water droplets. In order to determine the forced convective heat transfer coefficient at the inner boundary surface of vessel wall where the droplets impinge on and flow down, the transient temperatures of spray droplets when they reach the inner surface of the vessel wall after travelling from the spray nozzle through the pressurizer interior space occupied with the saturated steam-noncondensable hydrogen gas mixture have been predicted. The transient temperature distributions in the vessel wall have been obtained by using the finite element method, and the typical results have been provided. It has been shown that the results of thermal analysis are consistent with representation of the input transient and have plausible physical meaning.

• Nuclear Engineering and Technology
• /
• v.54 no.2
• /
• pp.608-616
• /
• 2022

Power peaking factors (PPF) is an important parameter for safe and efficient reactor operation. There are several methods to calculate the PPF at TRIGA research reactors such as MCNP and TRIGLAV codes. However, these methods are time-consuming and required high specifications of a computer system. To overcome these limitations, artificial intelligence was introduced for parameter prediction. Previous studies applied the neural network method to predict the PPF, but the publications using the ANFIS method are not well developed yet. In this paper, the prediction of PPF using the ANFIS was conducted. Two input variables, control rod position, and neutron flux were collected while the PPF was calculated using TRIGLAV code as the data output. These input-output datasets were used for ANFIS model generation, training, and testing. In this study, four ANFIS model with two types of input space partitioning methods shows good predictive performances with R² values in the range of 96%-97%, reveals the strong relationship between the predicted and actual PPF values. The RMSE calculated also near zero. From this statistical analysis, it is proven that the ANFIS could predict the PPF accurately and can be used as an alternative method to develop a real-time monitoring system at TRIGA research 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.

• Nuclear Engineering and Technology
• /
• v.40 no.6
• /
• pp.477-488
• /
• 2008

The objective of the paper is to analyze the thermally induced density wave oscillations in water cooled boiling water reactors. A transient thermal hydraulic model is developed with a characteristics-based implicit finite-difference scheme to solve the nonlinear mass, momentum and energy conservation equations in a time-domain. A two-phase flow was simulated with a one-dimensional homogeneous equilibrium model. The model treats the boundary conditions naturally and takes into account the compressibility effect of the two-phase flow. The axial variation of the heat flux profile can also be handled with the model. Unlike the method of characteristics analysis, the present numerical model is computationally inexpensive in terms of time and works in a Eulerian coordinate system without the loss of accuracy. The model was validated against available benchmarks. The model was extended for the purpose of studying the flow-induced density wave oscillations in forced circulation and natural circulation boiling water reactors. Various parametric studies were undertaken to evaluate the model's performance under different operating conditions. Marginal stability boundaries were drawn for type-I and type-II instabilities in a dimensionless parameter space. The significance of adiabatic riser sections in different boiling reactors was analyzed in detail. The effect of the axial heat flux profile was also investigated for different boiling reactors.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2003.06a
• /
• pp.1158-1161
• /
• 2003

For underwater milling of parts of nuclear reactor, a waterproof main spindle system was developed. which used a servo meter. Particularly, a waterproof system is available to cope with emergencies such as an electricity failure so that it prevents hazards from cutting radioactive materials. A developed spindle was designed to be capable of horizontal and vertical cutting and structural analysis was conducted with a FEM tool(Design Space) when the forces were loaded in each axis-direction.