• Proceedings of the KSR Conference
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• 2005.11a
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• pp.344-349
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• 2005

Cast steels have been mainly in rod type couplers which connect railway vehicles. These components, as important members, are exposed to repeat to repeated stresses and impact loading during stop and departure service for a long tie. The coupler suffers fatigue crack initiations and demage that cause need to repair weldment due to these loading conditions. Therefore, the heat affected zone of cast steel couple in rod type were evaluated in view of metallurgical weld characteristics after repair welding at laboratory. The specimens with two different welding techniques were evaluated after several welding conditions and post-heat treatments. Micro-vickers hardness and tensile tests and microstructural observations were conducted on heat affected zone of the weldment according to repair weld and post-heat treatment.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.100-101
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• 2018

• 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.

• International Journal of Safety
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• v.4 no.2
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• pp.6-11
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• 2005

The goals of the paper are to understand the impact damage behavior and identify the effect of surface protective materials on impact resistance in filament wound composite pressure vessels. For these, a series of low velocity impact tests was performed on specimens cutting from the full scale pressure vessel by the instrumented impact testing machine. The specimens are classified into two types, which are with and without surface protective material. The visualization for impact damage by two different impactors is made by metallurgical microscope. Based on the impact force history and damage, the impact resistance parameters were employed,rod its validity in identifying the damage resistance of filament wound composite pressure vessel was reviewed. As the results, the impact resistance of the filament wound composites and its dependency on the surface protective material were evaluated quantitatively

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 1999.08a
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• pp.172-179
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• 1999

As the martensite structure cause fracture failure during drawing from 5.5mm rod to 3.05mm dia. wire without additional heat treatment, the optium cooling condition to inhibit the occurrence of martensite was investigated. In order to get SAE9254+V quality, the effects of alloying element, vanadium on the mechanical properties were investigated. Based upon CCT and TTT curves and the results form cooling test in mill, optimun cooling was found in the condition of the laying head temp of 780$^{\circ}C$ and of the conveyor speed at 0.15m/sec with the whole cover closed. The wire rods produced under the condition showed the best mechanical properties of 120kg/$\textrm{mm}^2$ in TS and 50% in RA, having an excellent drawability. In vanadium added steels, tensile strength was improved without degrading elongation and charpy impact value. That means the strengthening by vanadium is mainly due to the grain refinement by the fine precipitates during tempering process.

• Journal of KSNVE
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• v.5 no.4
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• pp.515-524
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• 1995

Tubes in heat exchanger of fuel rods in reactor core are supported at intemediate point by support p0lates or springs. Current practice is, in case of heat exchanger, to allow clearance between tube and support plate for design and manufacturing consideration. And in case of fuel rod the clearance in support point can be generated due to the support spring force relaxation. Flow-induced vibration of a tube can cause it to impact or rub against support plate or against adjacent tubes and can result in fretting-wear. The tube-to- support dynamic interaction is used to relate experimental wear data from single-span test rigs to real multi-span heat exchanger configurations. The dynamic interaction cna be measured during experimental wear tests. However, the dynamic interaction is difficult to measure in real heat exchangers and, therefore, analytical techniques are required to estimate this interaction. This paper describels the nonlinear impact model of DAGS(Dynamic Analysis of Gapped Structure) code which simulates the tube response to external sinusodial or step excitation and predicts tube motion and tube-to-support dynamic interaction. Three experimental measurements-two single span rods excited by sinusodial force and a two span rod impacted by a steel ball are compared from the simulation nonlinear model of DAGS code. The simulation results from DAGS code are in good agreement with measurements. Therefore, the developed model of DAGS code is good analytical tool for estimating tube-to-support dynamic interaction in real heat exchangers.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1333-1344
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• 2019

An integrated approach of model simplification for high burnup spent nuclear fuel is proposed based on material calibration using optimization. The spent fuel rods are simplified into a beam with a homogenous isotropic material. The proposed approach of model simplification is applied to fuel rods with two kinds of interfacial configurations between the fuel pellets and cladding. The differences among the generated models and the effects of interfacial bonding efficiency are discussed. The strategy of model simplification adopted in this work is to force the simplified beam model of spent fuel rods to possess the same compliance and failure characteristics under critical loads as those that result in the failure of detailed fuel rod models. It is envisioned that the simplified model would enable the assessment of fuel rod failure through an assembly-level analysis, without resorting to a refined model for an individual fuel rod. The effective material properties of the simplified beam model were successfully identified using the integrated optimization process. The feasibility of using the developed simplified beam models in dynamic impact simulations for a horizontal drop condition is examined, and discussions are provided.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4166-4178
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• 2021

The Control Rod Drive Mechanism (CRDM) is an essential part of the reactor, which realizes the start-stop and power adjustment of the reactor by lifting and lowering the control rod assembly. As a moving part in CRDM, the latch directly contacts with the control rod assembly, and the life of latch is closely related to the service life of the reactor. In this paper, the relationship between the life of the latch and the step stress, friction stress, and impact stress in the process of movement is analyzed, and the optimization methodology and process of latch life based on the approximate model are proposed. The design variables that affect the life of the latch are studied through the experimental design, and the optimization objective of design variables based on the latch life is established. Based on this, an approximate model of the life of the latch is built, and the multi-objective optimization of the life of the latch is optimized through the NSGA-II algorithm.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.14 no.6
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• pp.512-519
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• 2004

This paper described the free vibration characteristics of Optimized H Type (OHT) spacer grids (SG) supporting the PWR fuel rod. The vibration test and the finite element (FE) analysis are performed under the free boundary condition and the clamped at two points (or three points) in the bottom which is the same one as the experimental condition for the dummy rod continuously supported by spacer grids. A modal test is conducted by the impulse excitation method using an impulse hammer and an accelerometer, and the TDAS module of the I-DEAS software is used to acquire and analyze the sensor signals. The softwares related to the FE analysis are the I-DEAS for the geometrical shape modeling and meshing, and the ABAQUS for solving. The fundamental frequency of the OHT SG by experiment under a clamped condition at two points is 175.18 Hz, and shows a bending mode. We think there is no resonance between the fuel rod and the SG because the SG's frequency is higher than that of the fuel rod existing in the range from 30 to 120 Hz. The fundamental frequency of the SG under the free boundary condition is 349.2 Hz showing a bending mode, and the results between the test and the analysis have a good agreement with maximum 7 ％ in error It is also found that the FE analysis model of the OHT SGs to analyze an impact, a buckling and vibration et al. has been generated with reliability.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2756-2766
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• 2024

Small rod-controlled pressurized water reactors (PWR) are the ideal energy source for vessel propulsion, benefiting from their high reactivity control efficiency. Since the control rods (CRs) increase the complexity of reactivity control, this paper seeks to study the performance of CRs in small rod-controlled PWRs to extend the lifetime and reduce power offset due to CRs. This study investigates CR grouping, move-in/out strategies, and axially non-uniform design effects on core neutron physics metrics. These metrics include axial offset (AO), core lifetime (CL), fuel utilization (FU), and radial power peaking factor (R-PPF). To simulate the movement of the CRs, a "Critical-CR-burnup" function was developed in OpenMC. In CR designs, the CRs are grouped into three banks to study the simultaneous and prioritized move-in/out strategies. The results show CL extension from 590 effective full power days (EFPDs) to 638-698 EFPDs. A lower-worth prioritized strategy minimizes AO and the extremum values decrease from -0.69 and + 0.81 to -0.28 and + 0.51. Although an axially non-uniform CR design can improve AO at the beginning of cycle (BOC), considering the overall CR worth change is crucial, as a significant decrease can adversely impact axial power distribution during the middle of cycle (MOC).