• 유체기계공업학회:학술대회논문집
• /
• 2005.12a
• /
• pp.240-243
• /
• 2005

This study deals with the shape optimization of a wire spacer fuel assembly of Liquid Metal Reactors (LMRs). The Response Surface based optimization Method is used as an optimization technique with the Reynolds-averaged Navier-Stokes analysis of fluid flow and heat transfer using Shear Stress Transport (SST) turbulence model as a turbulence closure. Two design variables namely, pitch to fuel rod diameter ratio and lead length to fuel rod diameter ratio are selected. The objective function is defined as a combination of the heat transfer rate and the inverse of friction loss with a weighting factor. Three level full-factorial method is used to determine the training points. In total, nine experiments have been performed numerically and the resulting datas have been analysed for optimization study. Also, a comparison has been made between the optimized surface and the reference one in this study.

• 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.36 no.2
• /
• pp.127-141
• /
• 2004

Various types of integrated steam generators, which integrate IHTS and a steam generator into a single unit of equipment for an LMR, were analyzed using an analytic solution with some simplification. The analysis showed that the undesirable reversed heat transfer, of which occurrence was previously observed only in an integrated single-region bundle type, can also occur in an integrated double-region bundle type. The mechanism of the reversed heat transfer occurrence in the double-region type is explained and it is shown the mechanism in the double-region type is completely different from that in the single-region type. Based on this finding, a method for preventing the aforementioned heat transfer is suggested. The performance of the four types of the integrated steam generators is assessed. For this assessment, a SG is actually designed for each type and the optimization in the geometric parameters and flow rate are optimized.

• Nuclear Engineering and Technology
• /
• v.35 no.4
• /
• pp.261-273
• /
• 2003

The characteristics of the recirculation operation of LMR which are different from the conventional plants such as PWR and fossil fuel plants were investigated using a computer code TSGS developed in this study. For simulating the transient behavior of the steam generation system, a water level control algorithm utilizing digital control hardware features was introduced. By investigation, the function of the recirculation operation was defined, the major features of the operation were found. Also good performance of the level control algorithm was confirmed.

• Proceedings of the KSME Conference
• /
• 2008.11b
• /
• pp.3175-3180
• /
• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Journal of Mechanical Science and Technology
• /
• v.20 no.12
• /
• pp.2061-2078
• /
• 2006

The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500$^{\circ}C$ and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.31 no.8
• /
• pp.663-671
• /
• 2007

In this work, shape optimization of a wire-wrapped fuel assembly in a liquid metal reactor has been carried out by combining a three-dimensional Reynolds-averaged Navier-Stokes analysis with the radial basis neural network method, a well known surrogate modeling technique for optimization. Sequential Quadratic Programming is used to search the optimal point from the constructed surrogate. Two geometric design variables are selected for the optimization and design space is sampled using Latin Hypercube Sampling. The optimization problem has been defined as a maximization of the objective function, which is as a linear combination of heat transfer and friction loss related terms with a weighing factor. The objective function value is more sensitive to the ratio of the wire spacer diameter to the fuel rod diameter than to the ratio of the wire wrap pitch to the fuel rod diameter. The optimal values of the design variables are obtained by varying the weighting factor.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
• /
• 2002.05b
• /
• pp.137-140
• /
• 2002

EM(ElectroMagnetic) pump is used for the purpose of transporting liquid sodium coolant with electrical conductivity in the LMR(Liquid Metal Reactor). In the present study, computer program for the pilot annular linear EM pump has been developed for the maximum flowrate with 200 l/min and maximum developing pressure with 3 bar. Firstly, Balance equation is induced by the equivalent circuit method which is commonly employed to analyze linear induction machines and the calculation of the hydraulic pressure drop. Then, design equation is converted to the computer program and optimum pump variables are determined by this code. The code is verified by the comparative analysis with the characteristic of the commercialized pump.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.26 no.1
• /
• pp.18-24
• /
• 2006

Reactor core and internal structures of a liquid metal reactor (LMR) can not be visually examined due to an opaque liquid sodium. The under-sodium viewing technique by using an ultrasonic wave should be applied far the visual inspection of reactor internals. In this study, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium viewing technique. The Lamb wave propagation of a waveguide sensor has been analyzed and the zero-order antisymmetric $A_0$ plate wave was selected as the application mode of the sensor. The $A_0$ plate wave can be propagated in the dispersive low frequency range by using a liquid wedge clamped to the waveguide. A new technique is presented which is capable of steering the radiation beam angle of a waveguide sensor without a mechanical movement of the sensor assembly The steering function of the ultrasonic radiation beam can be achieved by a frequency tuning method of the excitation pulse in the dispersive range of the $A_0$ mode. The technique provides an opportunity to overcome the scanning limitation of a waveguide sensor. The beam steering function has been evaluated by an experimental verification. The ultrasonic C-scanning experiments are performed in water and the feasibility of the ultrasonic waveguide sensor has been verified.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.26 no.5
• /
• pp.297-305
• /
• 2006

The thermal ratcheting deformation at the reactor baffle and upper internal structure of the liquid metal reactor (LMR) can occur due to movement of the hot sodium free surface. In in-service inspection of reactor internals of LMR, a new inspection technique should be developed for the detection of the thermal ratcheting damage. In this study, an inspection technique using ultrasonic guided wave is proposed for the detection of the thermal ratcheting damage of cylindrical vessels. A 316L stainless steel cylindrical shell specimen has been prepared. The thermal ratchet structural tests were cyclically performed by heat-up up to $550^{\circ}C$ with steep temperature gradients along the axial direction after cool-down by cooling water. Ultrasonic guided wave propagation has been characterized by analysis of dispersion curve of the stainless steel plate. The zero-order antisymmetric $A_0$ guided wave has been selected as the optimal mode for detection of the ratcheting deformation. It is confirmed that the thermal ratcheting deformation can be detected by the measurement of transit time difference of circumferentially propagated $A_0$ guided waves.