• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.201-202
• /
• 2004

• Nuclear Engineering and Technology
• /
• v.33 no.3
• /
• pp.283-297
• /
• 2001

The CHF in PWR rod bundles is usually predicted by the local flow correlation approach based on subchannel analysis while difficulty exists due to the existence of spacer grids especially with mixing vanes. In order to evaluate the effect of spacer grids on CHF, the experimental rod bundle data with various types of spacer grids were analyzed using the subchannel code, COBRA-IV-i. For the Plain grid data, a CHF correlation was described as a function of local flow conditions and heated length, and then the residuals of the CHF in mixing vaned grids predicted by the correlation were examined in various kinds of grids. In order to compensate for the residual, three parameters, distances between grids and from the last grids to the CHF site, and equivalent hydraulic diameter were introduced into a grid parameter function representing the remaining effect of spacer grids predicted most of the CHF data points in plaing grids within $\pm$20 percent error band. Good agreement with the CHF data was also shown when the grid parameter function for mixing vaned grids of a specific design was used to compensate for the residuals of the CHF data predicted by the correlation.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.502-507
• /
• 1998

Two different types the CANDU fuel bundles hue been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void paction distributions in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From calculated mixture enthalpy distribution at the exit of fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful assessing thermal behavior of the fuel bundle that could be used in CANDU reactors.

• Nuclear Engineering and Technology
• /
• v.36 no.2
• /
• pp.165-174
• /
• 2004

The distributed resistance model has been recently implemented into the MATRA-LMR code in order to improve its prediction capability over the wire-wrap model for a flow blockage analysis in the LMR. The code capability has been investigated using experimental data observed in the FFM (Fuel Failure Mock-up)-2A and 5B for two typical flow conditions in a blocked channel. The predicted results by the MATRA-LMR with a distributed resistance model agreed well with the experimental data for wire-wrapped subchannels. However, it is suggested that the parameter n in the distributed resistance model needs to be calibrated accurately for a reasonable prediction of the temperature field under a low flow condition. Finally, the analyses of a blockage for the assembly of the KALIMER design are performed. Satisfactory results by the MATRA-LMR code were obtained through and rerified a comparison with results of the SABRE code.

• Nuclear Engineering and Technology
• /
• v.55 no.12
• /
• pp.4357-4366
• /
• 2023

MARS-KS, a domestic regulatory confirmatory code of Republic of Korea, had been developed by integrating RELAP5/MOD2 and COBRA-TF. The integration of COBRA-TF allowed to extend the capability of MARS-KS, limited to one-dimensional analysis, to multi-dimensional analysis. The use of COBRA-TF was mainly focused on subchannel analyses for simulating multi-dimensional behavior within the reactor core. However, this feature has been remained as a legacy without ongoing maintenance. Meanwhile, MARS-KS also includes its own multidimensional component, namely MULTID, which is also feasible to simulate three-dimensional convection and diffusion. The MULTID is capable of modeling the turbulent diffusion using simple mixing length model. The implementation of the turbulent mixing is of importance for analyzing the reactor core where a disturbing cross-sectional structure of rod bundle makes the flow perturbation and corresponding mixing stronger. In addition, the presence of this turbulent behavior allows the secondary transports with net mass exchange between subchannels. However, a series of assessments performed in previous studies revealed that the turbulence model of the MULTID could not simulate the aforementioned effective mixing occurred in the subchannel-scale problems. This is obvious consequence since the physical models of the MULTID neglect the effect of mass transport and thereby, it cannot model the void drift effect and resulting phasic distribution within a bundle. Thus, in this study, the turbulence mixing model of the MULTID has been improved by means of the inter-channel mixing model, widely utilized in subchannel analysis, in order to extend the application of the MULTID to small-scale problems. A series of assessments has been performed against rod bundle experiments, namely GE 3X3 and PSBT, to evaluate the performance of the introduced mixing model. The assessment results revealed that the application of the inter-channel mixing model allowed to enhance the prediction of the MULTID in subchannel scale problems. In addition, it was indicated that the code could not predict appropriate phasic distribution in the rod bundle without the model. Considering that the proper prediction of the phasic distribution is important when considering pin-based and/or assembly-based expressions of the reactor core, the results of this study clearly indicate that the inter-channel mixing model is required for analyzing the rod bundle, appropriately.

• Nuclear Engineering and Technology
• /
• v.41 no.6
• /
• pp.797-806
• /
• 2009

The Multichannel Analyzer for Transient and steady-state in Rod Array - Liquid Metal Reactor for Flow Blockage analysis (MATRA-LMR-FB) code for the analysis of a subchannel blockage has been developed and evaluated through several experiments. The current version of the code is improved here by the implementation of a distributed resistance model which accurately considers the effect of flow resistance on wire spacers, by the addition of a turbulent mixing model, and by the application of a hybrid scheme for low flow regions. Validation calculations for the MATRA-LMR-FB code were performed for Oak Ridge National Laboratory (ORNL) 19-pin tests with wire spacers and Karlsruhe 169-pin tests with grid spacers. The analysis of the ORNL 19-pin tests conducted using the code reveals that the code has sufficient predictive accuracy, within a range of 5 $^{\circ}C$, for the experimental data with a blockage. As for the results of the analyses, the standard deviation for the Karlsruhe 169-pin tests, 0.316, was larger than the standard deviation for the ORNL 19-pin tests, 0.047.

• Nuclear Engineering and Technology
• /
• v.27 no.4
• /
• pp.518-531
• /
• 1995

A study has been Peformed to investigate the thermal margin increase by replacing the single-channel analysis model with a multichannel analysis model. h new critical heat flux(CHF) correlation, which is applicable to a 17$\times$17 Korean Fuel Assembly(KOFA)-loaded core, was developed on the basis of the local conditions predicted by the subchannel analysis code, TORC. The hot sub-channel analysis was carried out by using one-stage analysis methodology with a prescribed nodal layout of the core. The result of the analysis shooed that more than 5％ of the thermal margin can be recovered by introducing the TORC/KRB-1 system(multichannel analysis model) instead of the PUMA/ERB-2 system(single-channel anal)sis model). The thermal margin increase was attributed not only to the effect of the local thermal hydraulic conditions in the hot subchannel predicted by the code, but also to the effect of the characteristics of the CHF correlation.

• The Journal of Korean Institute of Electromagnetic Engineering and Science
• /
• v.14 no.12
• /
• pp.1300-1310
• /
• 2003

In this paper, the performance of a multicarrier CDMA system using adaptive modulation and adaptive subchannel allocation scheme is analyzed in Rayleigh fading channel. The proposed adaptive modulator consists of modulation schemes using QPSK 16 QAM, 64 QAM and 256 QAM and constellations are pointed by Gray code. In addition, the threshold of the analysis is average E$\_$b//N$\_$o/ when the BER is 1 %. In the multicarrier system with adaptive subchannel allocation scheme, each DS waveform of user is transmitted over the K subchannels with the biggest fading among L subchannels. In case of the proposed system, total 4 subchannels are used and data are transmitted over 2 subchannels with the biggest fading, which results in the threshold of each channels is 5.2 dB, 9 dB, 13.2 dB and 8.4 dB, 12.2 dB, 16.3 dB. In the case of proposed system, the BER of 10$\^$-3/ is satisfied if average E$\_$b//N$\_$o/ is 8.1 dB. This is increased performance of 12.9 dB in comparison with conventional system. The BPS according to average channel is needed average E$\_$b//N$\_$o/ of about 15 dB in 7 bit. In the case of subchannel error, the BER of 10$\^$-3/ is 13.6 dB and is declined about 5.5 dB.

• The Journal of Korean Institute of Communications and Information Sciences
• /
• v.30 no.4C
• /
• pp.200-206
• /
• 2005

Adaptive transmission techniques can improve the performance of wireless communication system by adaptively changing the transmission parameter such as modulation, code-rate, and power according to the channel state. For orthogonal frequency division multiplexing (OFDM) system, the adaptive transmission techniques can be applied to each subcarrier unit. In this paper, we consider the adaptive code-rate OFDM system in which optimal code-rate is applied to each subcarrier according to the subchannel state. Performance analysis show that $3\sim6$dB gain of SNR or up to $30\sim50\%$ increase of data rate are achieved in the condition of bit error rate $10^{-6}$.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.3154-3165
• /
• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.