정신과 폐쇄 병동 환경 평가에 영향을 미치는 임상변인 : 치료의 질적 향성을 위한 예비적 연구 The Preliminary Study for Promoting the Quality of Psychiatric Inpatient Treatment

서정석,류은정,이상미,한미희,최귀순,손인기,남범우 건국대학교 의과학연구소 2003 건국의과학학술지 Vol.13 No.-

Purpose: IS prelminary study was designed to assess the clinical variables influencing on the ward atmosphere in psychiatric closed ward and to promote the quality of psychiatric inpatient treatment. Method: 41 psychiatric inpatients were selected in Chung-ju hospital, Konkuk university. Ward Atmosphere Scale(W AS) was used to evaluate patients' perception of ward milieu. WAS score were compared among subgroups with respect to the clinical variables such as duration of admission, frequency of admission, and diagnosis. Results: Significantly, patients who have hospitalized for 1 month to 2 months reported lower score of WAS and patients who have hospitalized for longer than 2 months reported higher score of WAS. The relationship between educated levels and WAS score was not significant. First admitted patients reported significantly lower score of WAS. Bipolar patients reported significantly higher score of WAS and alcohol related patients reported significantly lower score of WAS. Conclusion: The 1st admitted patients, patients with alcohol problem and educated patients were had less satisfied with psychiatric closed ward. Thus, by considering these factors, more specified therapeutic approach and plan should be conducted.

고성능 표면탄성파 필터의 최적 설계에 관한 연구

신양호,김귀현,이진복,엄현석,이중근,박진석 漢陽大學校 工學技術硏究所 1999 工學技術論文集 Vol.8 No.1

본 논문에서는 고성능 표면탄성파 필터의 최적 설계를 위하여 다양한 IDT(interdigital transducer) 구조의 필터를 제작하고 각각의 주파수 응답특성을 비교 분석하였다. 기본 구조외에 MSC(micro strip-coupler), double split, reflector, ground line 등의 구조를 일반적으로 널리 사용되고 있는 Y-Z cut l-N(LiNbO3) 기판 위에 형성시켰다. 제작된 소자의 중심주파수, 통과 대역, 삽입손실등을 측정하였으며, 주파수 응답에 기초한 기존 모델의 효용성을 논의하였다. Effects of IDT structures on frequency responses of SAW filters are investigated to achieve a high performance SAW filter, Various types of IDTS, such as MSC(multistrip-coupler), double split, reflector, and ground line, are designed and patterned on a widely used Y-Z cut LN(LiNbO3) substrate. Center frequencies, band widths, and insertion losses are measured and compared. Invalidity of the conventional model based on the impulse response model is also discussed.

DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR

KWI-SEOK HA,HAE-YONG JEONG,WON-PYO CHANG,YOUNG-MIN KWON,CHUNGHO CHO,YONG-BUM LEE 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.6

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 ºC, 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.

COMPARISON OF THE DECAY HEAT REMOVAL SYSTEMS IN THE KALIMER-600 AND DSFR

KWI-SEOK HA,정해용 한국원자력학회 2012 Nuclear Engineering and Technology Vol.44 No.5

A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.

증례보고 : 말기 신부전증 환자의 동정맥루 수술에서 액와 상완 신경총 차단후 Lidocaine 정주로 촉발된 고칼륨성 심정지

송석영 ( Seok Young Song ),신흥동 ( Heung Dong Shin ),서귀주 ( Kwi Chu Seo ),정진용 ( Jin Yong Chung ),노운석 ( Woon Seok Roh ),김봉일 ( Bong Il Kim ) 대한마취과학회 2008 Korean Journal of Anesthesiology Vol.55 No.6

Axillary brachial plexus blockade (BPB) is commonly used as an anesthetic method for patients undergoing the creation of an arteriovenous fistula (AVF) during end-stage renal disease (ESRD). Several studies have shown that the combination of intravenous lidocaine and hyperkalemia in ESRD can produce severe conduction disturbance and asystole. Here, we report a case of cardiac arrest in a 41 year old male patient who manifested severe cardiac conduction disturbance during creation of an AVF. Sixty-five minutes after BPB, the intravenous therapeutic doses of lidocaine administered to treat frequent premature ventricular contractions aggravated his heart rhythm and produced a sine wave and ventricular fibrillation. It was assumed that ventricular fibrillation was induced by a combination of local anesthetics administered during BPB and systemic hyperkalemia as a result of the ESRD [ED highlight-please ensure my changes do not alter your intended meaning]. The patient was completely resuscitated 45 minutes after the cardiopulmonary resuscitation and correction of the hyperkalemia. (Korean J Anesthesiol 2008; 55: 756~60)

말기암 환자의 통증치료를 위한 지주막하강내 카테터 거치와 피하 이식형 약제 주입기 삽입

서귀주 ( Kwi Chu Seo ),정진용 ( Jin Yong Chung ),김호영 ( Ho Young Kim ),노운석 ( Woon Seok Rho ),김봉일 ( Bong Il Kim ),송석영 ( Seok Young Song ) 대한통증학회 2007 The Korean Journal of Pain Vol.20 No.2

DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR

Ha, Kwi-Seok,Jeong, Hae-Yong,Chang, Won-Pyo,Kwon, Young-Min,Cho, Chung-Ho,Lee, Yong-Bum Korean Nuclear Society 2009 Nuclear Engineering and Technology Vol.41 No.6

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.

Improvement of Liquid Droplet Entrainment Model in the COBRA-TF Code

Ha, Kwi-Seok,Jeong, Jae-Jun,Sim, Suk-Ku Korean Nuclear Society 1998 Nuclear Engineering and Technology Vol.30 No.3

The COBRA-TF liquid droplet entrainment models have been assessed and improved through various experiments. The COBRA-TF code uses the Wurtz entrainment model in the film mist flow regime and the mechanistic model based on the critical Weber number and critical vapor velocity in the hot wall flow regimes, respectively. The Wurtz model has been replaced with the modified Sugawara model. The assessment against the experiments by Hewitt, Keeys, Yanai, and Whalley showed the modified Sugawara model better predicts the steam-water as well as the air-water experiments for the film mist flow regime. For hot wall flow regime, the COBRA-TF entrainment model was modified using two methods, one with an increased critical Weber number and the other with the Yonomoto's critical vapor velocity model. The modified models were assessed using the FLECHT-SEASET bottom reflood tests. The results showed that the Yonomoto model best predicts the quenching time, whereas the local maximum rod temperature was not affected much.

Best estimate calculation and uncertainty quantification of sodium-cooled fast reactor using MARS-LMR code

Kang, Seok-Ju,Jeong, Hae-Yong,Bae, Sung-Won,Choi, Chi-Woong,Ha, Kwi-Seok,Suh, Jae-Seung Elsevier 2018 Annals of nuclear energy Vol.115 No.-

<P><B>Abstract</B></P> <P>The safety analysis of nuclear power plants has been performed using conservative approach based on conservative assumptions and boundary conditions to evaluate the safety margin of plant operation. However, this conservative approach could lead to unrealistic behavior predictions and eventually distort some phenomena in reactor systems. Therefore, the nuclear field moved towards an alternative best-estimate approach with uncertainty quantification in order to improve the phenomena prediction and to decrease the excessive conservatism in safety margins. In this study, the best estimate methodology is applied to improve the accuracy and reliability in safety analysis of an SFR. The applied best estimate methodology is based on the CSAU. This methodology is composed of three unique steps for evaluation of code capability, assessment and range of parameters, and sensitivity and uncertainty analysis. The primary purpose of this study is to evaluate the appropriateness of sensitivity parameters and its ranges, which have been determined through intensive experts’ panel discussion, by use of the data obtained from the EBR-II Unprotected Loss of Flow (ULOF) experiment. The MARS-LMR thermal–hydraulic code and the parallel computing platform integrated for uncertainty and sensitivity analysis (PAPIRUS) become the basic calculation tools in the study. Confirmation of data coverage is performed through the evaluation of coolant temperature in the instrumented subassemblies XX09. The appropriateness of parameters and its ranges are evaluated for three different cases: original parameters and ranges suggested in the MIRT, ±10% increased parameter ranges, and 200% increased axial reactivity feedback coefficient only. The case with the original parameters and ranges does not result in a valid data coverage, which means inadequate modeling accuracy for the ULOF scenario. The other two cases give complete coverage of EBR-II temperature data measured at the core top, which suggest the need of further refinement of reactivity models. The relative importance of the parameters is confirmed through the sensitivity analysis with respect to the Figures of Merit (FoM). The selected dominant parameters are the sodium density reactivity, above core load pad strain coefficient, core radial expansion reactivity coefficient and fuel axial expansion reactivity coefficient. The pump coastdown curve and the core inlet form loss are also found to be significant parameters during the transient.</P> <P><B>Highlights</B></P> <P> <UL> <LI> A best-estimate approach based on the CSAU is applied to the ULOF analysis of a pool-type SFR. </LI> <LI> The MARS-LMR code and the PAPIRUS tool are used to perform the uncertainty and sensitivity analyses. </LI> <LI> The sensitivity parameters and its ranges are determined through intensive experts’ panel discussion. </LI> <LI> The data coverage of code models is validated with the coolant temperature measured in EBR-II test. </LI> <LI> The relative importance of parameters is evaluated with the sensitivity study for the FoMs. </LI> </UL> </P>

COMPARISON OF THE DECAY HEAT REMOVAL SYSTEMS IN THE KALIMER-600 AND DSFR

Ha, Kwi-Seok,Jeong, Hae-Yong Korean Nuclear Society 2012 Nuclear Engineering and Technology Vol.44 No.5

A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.