• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.314-327
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• 1999

A subchannel analysis code MATRA applicable to PWRs and ALWRs has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-Rf-1. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and How distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. HATRA has been provided with an improved structure, various functions, and models to give more convenient user environment and to enhance the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the models for the lateral transport between adjacent subchannels have been improved to enhance the accuracy in predicting two-phase flow phenomena. The predictions of MATRA were compared with the experimental data on the flow and enthalpy distribution in some sample rod-bundle cases to evaluate the performance of MATRA. All the results revealed that the predictions of MATRA were better than those of COBRA-IV-I.

• Journal of Oral Medicine and Pain
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• v.34 no.3
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• pp.247-256
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• 2009

The purpose of this study was to investigate the changes of the pharyngeal space when the following appliances were inserted: the mandibular advancement appliance (MAA), tongue retaining appliance (TRA), and mandibular advancement-tongue retaining appliance (MATRA). Nine male dental students exhibiting Class I occlusion, normal body mass index (BMI), and no signs and symptoms of snoring were selected for this study. The three kinds of snoring appliances (MAA, TRA and MATRA) were fabricated for each subject. The mandibular advancement of the MAA and MATRA was set at a distance of 5 mm, and the TRA and MATRA were made to hold the tongue in front of the maxillary incisors by 10 to 20 mm. Lateral cephalometric radiographs of the following four states - with no appliance, MAA, TRA, and MATRA - were taken to examine any anatomical changes resulting from the application of the appliances. All four radiographs were traced and analyzed for twenty selected variables related to the pharyngeal space, cranio-cervical posture, and position of the soft palate and hyoid bone. According to the results of this study, there were significant increases in both the upper and lower oropharyngeal spaces when the mandible and tongue were protruded simultaneously, although there was a significant increase only in upper oropharyngeal space when the mandible or tongue was advanced separately. In conclusion, it is suggested that the MATRA may result in more positive effect on the control of snoring and OSA compared to a single use of the MAA or TRA, especially for the patients whose upper airway obstruction occurs in the lower oropharynx.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.678-683
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• 1998

액체금속로의 노심은 핵연료봉과 와이어랩에 의한 부수로로 구성된 복잡한 기하학적 구조의 집합체로 이루어져 있다. 이러한 액체금속로의 정상상태 및 과도상태 노심열수력 상세해석을 위하여 부수로 해석코드 MATRA-LMR 코드를 개발하고 있다. 본 논문에서는 ORNL 19 Pin 실험결과와 EBR-II 실험 모의시 정상상태 노심열수력 해석코드인 SLTHEN 코드 계산에 사용되었던 실험데이타를 사용하여 현재 MATRA-LMR 코드로 계산을 수행한 후 그 결과를 비교.분석함으로써, MATRA-LMR 코드의 개발 상태를 평가해 보았다 ORNL 19 Pin 실험과 MATRA-LMR 계산를 비교한 결과 실험을 정확히 예측하는 것으로 나타났다. SLTHEN 코드 계산결과와의 비교에서는 집합체 평균 출구온도와 부수로 최대 출구온도를 비교한 결과 두 코드의 계산은 약 3% 이하의 차이를 보이고 있다. 현재의 MATRA-LMR 코드는 단일 집합체 계산만 가능하나 앞으로의 작업을 통해 전 노심 해석이 가능하도록 다중 집합체 계산 코드로 개발할 예정이다.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1896-1906
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• 2020

The BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulation) benchmark calculations were performed by DeCART stand-alone and DeCART/MATRA multi-physics coupled code system to verify their accuracy. The solutions of DeCART stand-alone calculations for the control rod bank worth, detector signal, isothermal temperature coefficient, and critical boron concentration agreed very well with the measurements. The root-mean-square errors of the boron letdown curves for two-cycles were less than about 20 ppm, while the individual and total control rod bank worth agreed well within 7.3% and 2.4%, respectively. For the BEAVRS benchmark calculations at the beginning of burnup, the difference between DeCART simplified thermal-hydraulic stand-alone and DeCART/MATRA coupled calculations were not significantly large. Therefore, it is concluded that both the DeCART stand-alone code and the DeCART/MATRA multi-physics coupled code system have the capabilities to generate high fidelity transport solutions at core follow calculations.

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.295-306
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• 2009

The subchannel grade void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility were evaluated with the subchannel analysis code MATRA and the system code MARS. Fifteen test series from five different test bundles were selected for an analysis of the steady-state subchannel void distributions. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5% to 25%. The results of the transient calculations were also similar and were highly feasible. However, the computational aspects of the two codes were clearly different.

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

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.127.1-127.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.133-133
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• 1999

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

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.579-584
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• 1997

HP Workstation 및 IBM PC에서 사용 가능한 부수로 해석 코드 MATRA $\alpha$-버전을 개발하였다. 이 코드는 정확도 향상 및 사용자 편리를 위해 COBRA-IV-I코드에 비해 여러가지 기능들이 추가되었으며, 코드의 적용 범위를 신형원자로의 비정방형 집합체 노심에 확장할 수 있도록 압력손실 모형 등이 개선되었다. 또한 이상 유동장에서의 예측 정확도 향상을 위하여 부수로 잔의 횡방향 전달 모형을 개선하였다. 코드의 예측 성능을 평가하기 위해 세 중류의 집합체 유동분포 및 엔탈피 분포 실험 자료와 비교하였으며, 그 결과 기존의 COBRA-IV-I코드보다 향상된 결과를 보였다.