• The Journal of Korean Institute of Communications and Information Sciences
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• v.24 no.7A
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• pp.1005-1015
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• 1999

This paper is concerned with an improvement on IFFT (inverse fast Fourier transform) and FFT based baseband digital implementation algorithm for BPSK (binary phase shift keying)-modulated MC-CDMA (multicarrier-code division multiple access) systems, that is functionally equivalent to the conventional implementation algorithm, while reducing computational complexity and bandwidth requirement. We also derive an equalizer structure for the proposed implementation algorithm. The proposed algorithm is based on a variant of FFT algorithm that utilizes a N/2-point FFT/IFFT for simultaneous transformation and reconstruction of two N/2-point real signals. The computer simulations under additive white Gaussian noise channels and frequency selective fading channels using equal gain combiner and maximal ratio combiner diversities, demonstrate the performance of the proposed algorithm.

• 유체기계공업학회:학술대회논문집
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• 2004.12a
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• pp.180-186
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• 2004

Average Bi-directional flow tube was suggested to measure single and two phase flow rate. Its working principle is similar with Pilot tube, however, it makes it possible to eliminate the cooling system which is normally needed to prevent from flashing in the pressure impulse line of Pilot tube when it is used in the depressurization condition. 3-dimensional steady state flow analyses using FLUENT 5.4 code were performed to validate the application of the averagebi-directional flow tube in case of water and air flow In this study, sensitivity studies have been performed to optimize the design features of the average hi-directional flow tube which can be applied for the various experimental conditions. For Re numbers above 1000, the k values are nearly constant regardless of the Re numbers and flow types and calculation results and experimental data coincides quite well. The current FLUENT calculation results suggest that linearity of the k values in various design features of the average BDFT is highly promising, which means that it is quite reasonable to select the typical design of the average BDFT for the convenience of the experimental conditions.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.115-115
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• 1999

To demonstrate the applicability of RELAP5 to the prediction of the onset offlooding in the hot leg at the reflux condensation phase during mid-loop operation, numerical analysis is performed for the counter-current flow in a horizontal pipe with the inclined riser using the RELAP5/MOD3.2.2b code. It is found that the RELAP5, simulating the CCFL phenomena using interfacial friction along with the flow regime map in the horizontal pipe, produces unsatisfactory results. Under the CCFL condition, it is observed that large oscillation exists in the flow rate, void fraction, and etc. and the liquid flow rate is much lower than that predicted by the CCFL model measured in the experiment. The CCFL model of RELAP5 for the vertical volume is extended to the model for the horizontal and inclined volumes. The horizontal volume flow regime map and interfacial friction model coupled to the CCFL model are modified. And a new correlation developed from Kang's experiment is implemented to the CCFL model of RELAP5. With this modified RELAP5, the analysis of CCFL phenomena in the horizontal pipe and hot leg geometry is performed, and produces reasonable results in comparison with experimental data.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.567-578
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• 2015

In a nuclear reactor containment, wall condensation forms with noncondensable gases and their accumulation near the condensate film leads to a significant reduction in heat transfer. In the framework of nuclear reactor safety, the film condensation in the presence of noncondensable gases is of high relevance with regards to safety concerns as it is closely associated with peak pressure predictions for containment integrity and the performance of components installed for containment cooling in accident conditions. In the present study, CUPID code, which has been developed by KAERI for the analysis of transient two-phase flows in nuclear reactor components, is improved for simulating film condensation in the presence of noncondensable gases. In order to evaluate the condensate heat transfer accurately in a large system using the two-fluid model, a mass diffusion model, a liquid film model, and a wall film condensation model were implemented into CUPID. For the condensation simulation, a wall function approach with a heat/mass transfer analogy was applied in order to save computational time without considerable refinement for the boundary layer. This paper presents the implemented wall film condensation model, and then introduces the simulation result using the improved CUPID for a conceptual condensation problem in a large system.

• Journal of the Korea Society for Simulation
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• v.13 no.2
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• pp.13-21
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• 2004

The thermal-hydraulic model ARTS which was based on the RETRAN-3D code adopted in the domestic full-scope power plant simulator which was provided in 1998 by KEPRI. Since ARTS is a generalized code to model the components with control volumes, the smaller time-step size should be used even if converged solution could not get in a single volume. Therefore, dedicated models which do not force to reduce the time-step size are sometimes more suitable in terms of a real-time calculation and robustness. In the case of PRT(Pressurizer Relief Tank) model, it is consist of subcooled water in bottom and non-condensable gas in top. The sparger merged under subcooled water enhances condensation. The complicated thermal-hydraulic phenomena such as condensation, phase separation with existence of non-condensable gas makes difficult to simulate. Therefore, the PRT volume can limit the time-step size if we model it with a general control volume. To prevent the time-step size reduction due to convergence failure for simulating this component, we developed a dedicated model for PRT. The dedicated model was expected to provide substantially more accurate predictions in the analysis of the system transients. The results were resonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with the ANSI/ANS-3.5-1998 simulator software performance criteria and RETRAN-3D results.

• Nuclear Engineering and Technology
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• v.32 no.6
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• pp.618-635
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• 2000

A total of 17 experimental data for the onset of slugging, which is assumed to be the precursor of the condensation-induced waterhammer (CIWH), have been obtained for various How rates of water Incorporating the most recent correlations of interfacial heat transfer and friction factor developed for a circular geometry and using an improved criterion of transition from stratified to a slug flow, two existing analytical models to predict lower and upper bounds for CIWH have been upgraded. Applicability of the present as well as existing CIWH models has been tested by comparison with two sets of CIWH data. The result of this comparison shows that the applicability of the present as well as existing models is reasonably good. Based on the present models for CIWH, a computer code entitled as“KAIST-CIWH”has been developed and sample guide charts to find CIWH free regions for a given combination of major flow parameters in a long horizontal pipe have been presented along with the results of parametric studies of major parameters (D, P, $T_{f,in}$, and L/D) on the critical inlet water flow rate（$W_{f,in}_crit$ for both lower and upper bounds. In addition, two simple formulas for lower and upper bounds that can be used in an emergency for quick results have been presented.

• The KIPS Transactions:PartD
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• v.14D no.7
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• pp.753-762
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• 2007

Traceability has been held as an important factor in testing activities as well as model driven development. Vertical traceability affords us opportunities to improve manageability from models and test cases to code in testing and debugging phase. Traceability also makes overcome to difficulties of going up-and-down abstraction level to find out error spot of faults discovered by testing This paper represents a vertical test method which connects a system test level and an integration test level in a test stage by using UML. Experiment of how traceability works and how effective focus on error spots has been included using concrete examples of tracing from models to the code.

• Nuclear Engineering and Technology
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• v.19 no.3
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• pp.198-204
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• 1987

A plant simulation code, MCSIM (Micro-Computer SIMulator), has been developed to simulate plant transient accidents for pressurized water reactors. Reactor coolant system is modeled using decoupled energy and momentum equations, drift flux two-phase flow model and integral momentum equation. A two-fluid pressurizer model is used to simulate the pressurizer dynamics. Pot Boiler model is used for steam generator, steady-state decoupled energy and momentum equations for secondary side system, and point kinetics equations for nuclear power calculation. For test of the present version of MCSIM, complete loss of flow and RCCA withdrawal accidents are calculated with MCSIM. The results are compared with those in FSAR of KNU 5 & 6.

• Nuclear Engineering and Technology
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• v.36 no.5
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• pp.388-402
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• 2004

A phenomena identification and ranking table(PIRT) was developed for a main steam line break (MSLB) event for the Advanced Power Reactor-1400 (APR-1400). The selectee event was a double-ended steam line break at full power, with the reactor coolant pump running. The developmental panel selected the fuel performance as the primary safety criterion during the ranking process. The plant design data, the results of the APR-1400 safety analysis, and the results of an additional best-estimate analysis by the MARS computer code were used in the development of the PIRT. The period of the transient was composed of three phases: pre-trip, rapid cool-down, and safety injection. Based on the relative importance to the primary evaluation criterion, the ranking of each system, component, and phenomenon/process was performed for each time phase. Finally, the knowledge-level for each important process for certain components was ranked in terms of existing knowledge. The PIRT can be used as a guide for planning cost-effective experimental programs and for code development efforts, especially for the quantification of those processes and/or phenomena that are highly important, but not well understood.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.466-475
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• 2017

Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal-hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.