국외 열수력 해석 코드 개발 동향

정법동,정재준 한국원자력산업회의 2002 원자력산업 Vol.22 No.12

Improvements to the RELAP5/MOD3 Reflood Model and Assessment

정법동,이영진,박찬억,최철진,황태석,Chung, B.D.,Lee, Y.J.,Park, C.E.,Choi, C.J.,Hwang, T.S. Korean Nuclear Society 1994 Nuclear Engineering and Technology Vol.26 No.2

Several improvements to the RELAP5/MOD3 reflood model hate been made. These improvement were made to correct deficiencies in the reflood model identified by the assessment of the RELAP5/MOD3 code against FLECHT-SEASET experiments. The improvements consist of modification of reflood wall heat transfer package and adjusting the droplet size in dispersed flow regime. The time smoothing of wall vaporization and level tracking of transition flow are also added to eliminate the pressure spikes and level oscillation during reflood process. Assessment of the improved model against FLECHT-SEASET experimental data and application of LBLOCA analysis for plant shows that the deficiencies have been corrected. FLECHT-SEASET 실험에 대한 REIAP5/MOD3 평가시에 밝혀진 코드결함을 수정하기 위하여 RELAP5/MOD3 재관수 모델을 개선하였다. 모델개선은 재관수 열전달 모델의 수정과 분산유동영역의 액적 크기의 조절을 통하여 이루어졌으며 재관수 계산시 발생되는 압력 spike와 수위진동 등의 결함을 개선하기 위하여 벽면비등모델의 time-smoothing과 천이 유동시의 level tracking모델도 첨가되었다. FLECHT-SEASET 실험에 대한 개선모델의 검증과 발전소의 대형냉각재 상실 사고해석 응용에서 코드결함이 개선되었음을 알 수 있었다.

원자력 계통 열수력 해석코드의 3차원 모듈을 이용한 LOFT L2-5 실험 분석

김준혁,정재준,정법동 대한기계학회 2019 大韓機械學會論文集B Vol.43 No.8

A multi-dimensional approach is necessary for a better understanding of complex thermal-hydraulic phenomena in a reactor pressure vessel (RPV) during a large-break loss of coolant accident (LBLOCA). LBLOCA is one of the design basis accidents in a pressurized water reactor. The thermal-hydraulic system analysis code, MARS-KS, has a multi-dimensional module which enables both one-dimensional and three-dimensional analyses. However, the multi-dimensional module has not been studied extensively. In this study, the result of a LOFT L2-5 LBLOCA experiment was analyzed using the MARS-KS code. The multi-dimensional and one-dimensional modules were respectively used for the multi-dimensional phenomena in RPV and for the remaining regions. The results of the three-dimensional analysis for the RPV thermalhydraulics were compared with the experimental data for the assessment of the multi-dimensional module. 가압경수형 원자력발전소의 설계기준사고 중 하나인 대형냉각재상실사고(LBLOCA: Large-Break Loss of Coolant Accident) 발생 시 원자로 압력용기 내에서 발생하는 복잡한 열수력 현상을 보다 잘 이해하기 위해 다차원 해석이 필요하다. 열수력계통해석 코드인 MARS-KS는 다차원(multi-d) 모듈을 포함하고 있어 1차원 해석뿐만 아니라 3차원 해석도 가능하지만, 다차원 모듈에 대한 평가 계산이 부족한 실정이다. 본 연구에서는 대표적인 LBLOCA 실험인 LOFT L2-5 실험의 원자로 압력용기(RPV: Reactor Pressure Vessel)에서 발생하는 다차원 열수력 현상을 3차원 모듈로 분석하였다. 원자로 압력용기를 제외한 나머지 부분은 1차원으로 해석하였다. RPV 내의 온도 분포 3차원 해석 결과를 실험 결과와 비교함으로써 MARS-KS 코드의 다차원 모듈을 이용해 RPV를 모델링 하였을 때 LBLOCA 동안 RPV 내의 열수력학적 거동을 더욱 정확하게 분석할 수 있다는 것을 확인하였다.

Quantification of Realistic Discharge Coefficients for the Critical Flow Model of RELAP5/MOD3/KAERl

권태순,정법동,이원재,이남호,허재영,Kwon, T.S.,Chung, B.D.,Lee, W.J.,Lee, N.H.,Huh, J.Y. Korean Nuclear Society 1995 Nuclear Engineering and Technology Vol.27 No.5

RELAP5 /MOD3/KAERl의 임계유동모델을 위한 실제적인 배출계수들을 9개의 MARVIKEN 임계유동실험 의 평가계산을 통하여 과냉각과 이상임계유동에 대하여 구하였다. 선택된 실험에는 높은 초기 과냉각도와 큰 노즐 세 장비(L/D)인 것들이 포함되었다. 코드의 평가결과는 RELAP5/MOD3/KAERI은 과냉각임계유동을 크게 예측하고 이 상임계유동은 작게 예측함을 보이고 있다. 이러한 결과들을 이용하여 임계유동모델의 실제적인 배출계수들을 반복법으로 정량화 하였다. 실제적인 배출계 수는 과냉각임계유동이 0.89 그리고 이상임계유동이 1.07로 결정되었으며 관련 표준편차는 각 각 0.0349과 0.1189이다. 본 연구로부터 얻어진 결과는 대형냉각재 상실사고의 실제적인 계통반응 계산과 비상노심냉각계통 성능평가에 적용할 수 있다. The realistic discharge coefficient for the critical How model of RELAP5/AOD3/KAERI are determined for the subcooled and too-phase critical flow by assessments of nine MARVIKEN Critical flew Test(CFT). The selected test runs include a high initial subcooling and large nozzle aspect rat-io(L/D). The code assessment results show that RELAP5/MOD3/KAERI over-predicts the subcooled critical flow and under-predicts the two-phase critical flow. Using these result, the realistic discharge coefficients of critical flow models are quantified by an iterative method. The realistic discharge coefficients are determined to be 0.89 for the subcooled critical How and 1.07 for the two-phase critical flow, and the associated standard deviations are 0.0349 and 0.1189, respectively. The results obtained from this study can be applied to calculate the realistic system response of Large Break Loss of Coolant Accident and to evaluate the realistic Emergency Core Cooling System performance.

발전소 계통해석을 위한 MARS 코드의 다차원 이상 난류 유동 모델 검증계산

이석민,정법동,이은철,배성원 한국에너지학회 2006 에너지공학 Vol.15 No.1

원자력발전소의 다차원 이상 유동 현상을 적절히 모사하기 위해 일차원 계통해석 코드에 삼차원 유동모델을 적용하였다. 그 중 다차원모델에 새롭게 적용된 이상 난류모델을 검증하기 위해 사각 slab 내부의 단상유동을 계산하여 상용 CFD 코드의 계산결과와 비교하였다. 그 결과 단상유동의 경우 난류 모델의 계산이 적절히 수행됨을 알 수 있었다. 또한 다차원 이상 유동 계산을 검증하기 위해서 RPI에서 수행된 물-공기 다차원 실험의 기포율 분포를 비교하였다. 그 결과 다차원 모델의 이상 유동 계산을 위해서는 일차원 기반의 유동양상 맵 중 수평 분리 유동양상이 제거 되어야 함을 알 수 있었다. 이와 같이 유동양상 맵을 수정하여 모사한 계산결과가 실험에서 측정된 기포율의 경향을 잘 따르는 것으로 계산되었다.

Analysis of Uncertainty Quantification Method by Comparing Monte-Carlo Method and Wilks' Formula

이성욱,정법동,방영석,배성원 한국원자력학회 2014 Nuclear Engineering and Technology Vol.46 No.4

An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performedand compared with the tolerance level determined by the Wilks’ formula. The uncertainty range and distribution of eachinput parameter associated with the LOCA phenomena were determined based on previous PIRT results and documentationduring the BEMUSE project. Calulations were conducted on 3,500 cases within a 2-week CPU time on a 14-PC clustersystem. The Monte-Carlo exercise shows that the 95% upper limit PCT value can be obtained well, with a 95% confidencelevel using the Wilks’ formula, although we have to endure a 5% risk of PCT under-prediction. The results also show thatthe statistical fluctuation of the limit value using Wilks’ first-order is as large as the uncertainty value itself. It is thereforedesirable to increase the order of the Wilks’ formula to be higher than the second-order to estimate the reliable safety marginof the design features. It is also shown that, with its ever increasing computational capability, the Monte-Carlo method isaccessible for a nuclear power plant safety analysis within a realistic time frame.

Phenomena Identification and Ranking Table for the APR-1400 Main Steam Line Break

송진호,방영석,정법동,정재준,백원필 한국원자력학회 2004 Nuclear Engineering and Technology Vol.36 No.5

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 selected 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.

원자력 안전과 경제성 향상을 위한 열유동 실험 및 해석 기술 개발

백원필,송철화,천세영,정법동,정문기,김경두,정흥준,김희동 한국원자력산업회의 2005 원자력산업 Vol.25 No.8

노외 중대사고 분석을 위한 격납건물 열수력 솔버 예비 개발

한금호(Kum Ho Han),정법동(Bub-Dong Chung),추연준(Yeon-Jun Choo) 한국유체기계학회 2023 한국유체기계학회 학술대회 논문집 Vol.2023 No.7

Moving reactor model for the MULTID components of the system thermal-hydraulic analysis code MARS-KS

서형주,최문희,박상욱,김건우,조형규,정법동 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.11

Marine reactor systems experience platform movement, and therefore, the system thermal-hydraulic analysis code needs to reflect the motion effect on the fluid to evaluate reactor safety. A moving reactor model for MARS-KS was developed to simulate the hydrodynamic phenomena in the reactor under motion conditions; however, its applicability does not cover the MULTID component used in multidimensional flow analyses. In this study, a moving reactor model is implemented for the MULTID component to address the importance of multidimensional flow effects under dynamic motion. The concept of the volume connection is generalized to facilitate the handling of the junction of MULTID. Further, the accuracy in calculating the pressure head between volumes is enhanced to precisely evaluate the additional body force. Finally, the Coriolis force is modeled in the momentum equations in an acceleration form. The improvements are verified with conceptual problems; the modified model shows good agreement with the analytical solutions and the computational fluid dynamic (CFD) simulation results. Moreover, a simplified gravity-driven injection is simulated, and the model is validated against a ship flooding experiment. Throughout the verifications and validations, the model showed that the modification was well implemented to determine the capability of multidimensional flow analysis under ocean conditions.