• Journal of Korean Association for Spatial Structures
• /
• v.14 no.4
• /
• pp.25-30
• /
• 2014

• Nuclear Engineering and Technology
• /
• v.38 no.3
• /
• pp.293-300
• /
• 2006

A liquid metal reactor (LMR) operated at high temperatures is subjected to both cyclic mechanical loading and thermal loading; thus, creep-fatigue is a major concern to be addressed with regard to maintaining structural integrity. The Korea Advanced Liquid Metal Reactor (KALIMER), which has a normal operating temperature of $545^{\circ}C$ and a total service life time of 60 years, is composed of various cylindrical structures, such as the reactor vessel and the reactor baffle. This study focuses on the creepfatigue crack initiation for a cylindrical Y-junction structure made of 316 stainless steel (SS), which is subjected to cyclic axial tensile loading and thermal loading at a high-temperature hold time of $545^{\circ}C$. The evaluation of the considered creep-fatigue crack initiation was carried out utilizing the ${\sigma}_d$ approach of the RCC-MR A16 guide, which is the high-temperature defect assessment procedure. This procedure is based on the total accumulated strain during the service time. To confirm the evaluated result, a high-temperature creep-fatigue structural test was performed. The test model had a circumferential through wall defect at the center of the model. The defect front of the test model was investigated after the $100^{th}$ cycle of the testing by utilizing a metallurgical inspection technique with an optical microscope, after which the test result was compared with the evaluation result. This study shows how creep-fatigue crack initiation for a high-temperature structure can be predicted with conservatism per the RCC-MR A16 guide.

• Nuclear Engineering and Technology
• /
• v.53 no.5
• /
• pp.1556-1568
• /
• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect stress conditions, mechanical behaviors and thermal-hydraulic performance of the fuel assembly. This paper is the Part II work of a two-part study devoted to analyzing the complex unique mechanical deformation and thermal-hydraulic characteristics for the typical plate-type fuel assembly under irradiation effect, which is on the basis of developed and verified numerical thermal-fluid-structure coupling methodology under irradiation in Part I of this work. The mechanical deformation, thermal-hydraulic performance and Mises stress have been analyzed for the typical plate-type fuel assembly consisting of support plates under non-uniform irradiation. It was interesting to observe that: the plate-type fuel assembly including the fuel plates and support plates tended to bend towards the location with maximum fission rate; the hot spots in the fuel foil appeared at the location with maximum thickness increment; the maximum Mises stress of fuel foil was located at the adjacent location with the maximum plate thickness increment et al.

• Nuclear Engineering and Technology
• /
• v.49 no.7
• /
• pp.1513-1523
• /
• 2017

Improvement of numerical analysis methods has been required to solve complicated phenomena that occur in nuclear facilities. Particularly, fluid-structure interaction (FSI) behavior should be resolved for accurate design and evaluation of complex reactor vessel internals (RVIs) submerged in coolant. In this study, the FSI effect on dynamic characteristics of RVIs in a typical 1,000 MWe nuclear power plant was investigated. Modal analyses of an integrated assembly were conducted by employing the fluid-structure (F-S) model as well as the traditional added-mass model. Subsequently, structural analyses were carried out using design response spectra combined with modal analysis data. Analysis results from the F-S model led to reductions of both frequency and Tresca stress compared to those values obtained using the added-mass model. Validation of the analysis method with the FSI model was also performed, from which the interface between the upper guide structure plate and the core shroud assembly lug was defined as the critical location of the typical RVIs, while all the relevant stress intensities satisfied the acceptance criteria.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2008.11a
• /
• pp.687-692
• /
• 2008

In a nuclear power plant, loose part monitoring and its diagnostic technique is one of the major issues for ensuring the structural integrity of the reactor system. Typically, accelerometers are mounted on the surface of a reactor vessel to localize impact location caused by the impact of metallic substances on the reactor system. However, in some cases, the number of the accelerometers is not enough to estimate the impact location precisely. In such a case, one of alternative plan is to utilize another type sensors that can measure the vibration of the reactor structure even though the measuring frequency ranges are different from each others. The AE sensors installed on the reactor structure can be utilized as additional sensors for loose part monitoring. In this paper, we proposed a new method to estimate impact location by using both accelerometer signal and AE signal, simultaneously. The feasibility of the proposed method is verified by an experiment. The experimental results demonstrate that we can enhance the reliability and precision of the loose part monitoring.

• Progress in Superconductivity and Cryogenics
• /
• v.18 no.2
• /
• pp.18-20
• /
• 2016

The study of superconducting fault current limiter (SFCL) is continuously being studied as a countermeasure for reducing fault-current in the power system. When the fault occurred in the power system, the fault-current was limited by the generated impedance of SFCLs. The operational characteristics of the flux-offset type SFCL according to turn ratios between the primary and the secondary winding of a reactor were compared in this study. We connected the secondary core to a superconductor and a SCR switch in series in the suggested structure. The fault current in the primary and the secondary winding of the reactor and the voltage of the superconductor on the secondary were measured and compared. The results showed that the fault current in the load line was the lowest and the voltage applied at both ends of the superconductor was also low when the secondary winding of the reactor had lower turn ratio than the primary. It was confirmed based on these results that the turn ratio of the secondary winding of the reactor must be designed to be lower than that of the primary winding to reduce the burden of the superconductor and to lower the fault current. Also, the suggested structure could increase the duration of the limited current by limiting the continuous current after the first half cycle from the fault with the fault current limiter.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.18 no.11
• /
• pp.1143-1149
• /
• 2008

In a nuclear power plant, loose part monitoring and its diagnostic technique is one of the major issues for ensuring the structural integrity of the reactor system. Typically, accelerometers are mounted on the surface of a reactor vessel to localize impact location cavsed by the impact of metallic substances on the reactor system. However, in some cases, the number of the accelerometers is not enough to estimate the impact location precisely. In such a case, one of alternative plan is to utilize another type sensors that can measure the vibration of the reactor structure even though the measuring frequency ranges are different from each others. The AE sensors installed on the reactor structure can be utilized as additional sensors for loose part monitoring. In this paper, we proposed a new method to estimate impact location by using both accelerometer signal and AE signal, simultaneously. The feasibility of the proposed method is verified by an experiment. The experimental results demonstrate that we can enhance the reliability and precision of the loose part monitoring.

• Wind and Structures
• /
• v.23 no.5
• /
• pp.421-447
• /
• 2016

Wind-induced and earthquake-induced excitations on tall structures can be effectively controlled by Tuned Liquid Damper (TLD). This work presents a numerical simulation procedure to study the performance of tuned liquid tank- structure system through ${\sigma}$-transformation based fluid-structure coupled solver. For this, a 'C' based computational code is developed. Structural equations are coupled with fluid equations in order to achieve the transfer of sloshing forces to structure for damping. Structural equations are solved by fourth order Runge-Kutta method while fluid equations are solved using finite difference based sigma transformed algorithm. Code is validated with previously published results. The minimum displacement of structure is observed when the resonance condition of the coupled system is satisfied through proper tuning of TLD. Since real-time excitations are random in nature, the performance study of TLD under random excitation is also carried out in which the Bretschneider spectrum is used to generate the random input wave.

• Nuclear Engineering and Technology
• /
• v.52 no.10
• /
• pp.2204-2220
• /
• 2020

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding. Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant. In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool. In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 ℃ higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.

• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 1993.04a
• /
• pp.165-172
• /
• 1993

A non-linear mathematical model has been developed for the dynamic analysis of the reactor internals. The model includes a lumped mass and stiffness with non-linear members such as gap-spring. As hydrodynamic couplings have also been considered in the model, the effect of fluid/structure interaction between internals components due to their immersion in a confining fluid can be studied for the dynamic response analysis. The reactor internals responses for seismic and pipe break excitations have been calculated for the case of with-and without-hydrodynamic couplings.