열분배모델을 이용한 수직유로에서의 저압 미포화비등 해석

이바로(Ba-Ro Lee),이연건(Yeon-Gun Lee) 대한기계학회 2016 大韓機械學會論文集B Vol.40 No.7

벽면비등 모델로 열분배모델을 채택하는 CFD 스케일의 전산해석코드는 저압 조건에서 미포화비등 발생 시 2상유동 변수의 해석 정확도가 낮은 것으로 알려진다. 본 연구에서는 열분배모델을 기반으로 벽면비등 현상을 예측하는 열수력 기기해석코드인 CUPID 코드를 이용하여 수직상향류 미포화비등 실험을 해석하였다. 10 bar 이상의 고압 조건에서는 CUPID 코드의 기포율 예측 정확도가 높았으나, 대기압 주변의 저압 조건에서는 기포율 분포에 대한 해석결과가 실험결과와 큰 차이를 보였다. 따라서 열분배모델 내 주요 인자에 사용되는 부모델에 대한 민감도 분석을 수행하였으며, 저압 조건 미포화비등 예측에 적합한 최적 부모델 조합을 선정하였다. 또한, 열분배모델 내 주요 인자 중 하나인 K-인자가 기포율에 미치는 영향을 평가하였다. Most CFD codes, that mainly adopt the heat partitioning model as the wall boiling model, have shown low accuracies in predicting the two-phase flow parameters of subcooled boiling phenomena under low pressure conditions. In this study, a number of subcooled boiling experiments in vertical channels were analyzed using a thermal-hydraulic component code, CUPID. The prediction of the void fraction distribution using the CUPID code agreed well with experimental data at high-pressure conditions; whereas at low-pressure conditions, the predicted void fraction deviated considerably from measured ones. Sensitivity tests were performed on the submodels for major parameters in the heat partitioning model to find the optimized sets of empirical correlations suitable for low-pressure subcooled flow boiling. The effect of the K-factor on the void fraction distribution was also evaluated.

CUPID 코드를 이용한 수직 증기 제트에 의한 헬륨 성층 침식 해석

손지현(Ji Hyun Sohn),조윤제(Yun Je Cho),윤한영(Han Young Yoon) 대한기계학회 2020 大韓機械學會論文集B Vol.44 No.7

가압 경수로에서 사고 시에 격납 건물 내부로 방출된 수소는 폭발의 잠재성을 가지고 있기 때문에 격납 건물의 구조적 건전성에 위협이 될 수 있다. 따라서 수소의 폭발 위험을 줄이기 위한 수소 농도의 저감 방안이 필요하며, 원자로 냉각 계통 파단부에서 방출되는 증기 제트에 의한 수소 성층 침식은 대기 혼합을 통해 수소의 농도를 저감시킬 수 있는 중요한 현상 중의 하나이다. 본 연구에서는 3차원 CFD 코드인 CUPID를 이용한 격납 건물 수소 분포 예측과 관련하여, 제트에 의한 성층 침식 조건에 대해 해석에 적용되는 난류 모델과 복사 열전달 모델이 기체 농도 변화 및 온도 분포 예측에 미치는 영향을 평가하기 위한 민감도 계산을 수행하였고, 이를 바탕으로 유동 장애물이 고려되는 조건에서의 제트-성층 침식 현상에 대한 코드 검증 계산을 수행하였다. 코드 검증에는 HYMERES-2 프로젝트에서 수행된 수직으로 주입되는 증기 제트에 의한 헬륨 성층 침식 실험이 이용되었다. Hydrogen released into containment during an accident in a pressurized water reactor has the potential for explosion, which can be a threat to the structural integrity of containment. Therefore, it is necessary to mitigate the concentration of hydrogen to reduce the risk of explosion. Hydrogen stratification erosion by steam jet injected from a broken part of the reactor coolant system is one of the important phenomena that can mitigate the concentration of hydrogen through atmospheric mixing. The three-dimensional computational fluid dynamics analysis code CUPID was used in this study. A sensitivity analysis was conducted on the turbulence and radiative heat transfer models implemented in CUPID. For the validation of the analysis, the experimental data of HYMERES-2 were used. HYMERES-2 is a joint international project to investigate the helium stratification erosion phenomena by a vertical steam jet.

CUPID코드의 유체 물성치 변화를 고려한 자연대류 해석

이승준(S.J. Lee),박익규(I.K. Park),윤한영(H.Y. Yoon),김정우(J. Kim) 한국전산유체공학회 2015 한국전산유체공학회지 Vol.20 No.4

Without electirc power to cool down the hot reactor core, passive systems utilizing natural circulation are becoming a big specialty of recent neculear systems after the severe accident in Fukusima. When we consider the natural circulation in a pool, thermal mixing phenomena may start from single phase circulation and can continue to two phase condition. Since the CUPID code, which has been developed for two-phase flow analysis, can deal with the phase transition phenomena, the CUPID would be pertinent to natural convection problems in single- and two-phase conditions. Thus, the CUPID should be validated against single- and two-phase natural circulation phenomena. For the first step of the validation process, this study is focused on the validation of single-phase natural circulation. Moreover, the CUPID code solves the fluid properties by the relationship to pressure and temperature from the steam table considering non-condensable gas effects, so that the effects from variable properties are included. Simple square thermal cavity problems are tested for laminar and turbulent conditions against numerical and experimental data. Throughout the investigation, it is found that the variable properties can affect the flow field in laminar condition, but the effect becomes weak in turbulence condition, and the CUPID code implementing steam table is capable of analyzing single phase natural circualtion phenomena.

CUPID 코드와 MARS 코드를 이용한 기기/계통 다중스케일 연계 해석 코드 구현

박익규(I.K. Park) 한국전산유체공학회 2016 한국전산유체공학회지 Vol.21 No.3

In this study, direct code coupling, in which two codes share a single flow field, was conducted using 3-dimensional high resolution thermal hydraulics code, CUPID and 1-dimensional system analysis code, MARS. This approach provide the merit to use versatile capability of MARS for nuclear power plants and 3-dimensional T/H analysis capability of CUPID. Numerical Method to directly couple CUPID and MARS was described in this paper. The straight flow and manometer flow oscillation were calculated to verify conservation of coupled CUPID/MARS code in mass, momentum, and energy. This verification calculations indicates that the CUPID/MARS is coupled appropriately in numerical aspect and the coupled code can be applied to nuclear reactor thermal hydraulics after validation against integral transient experiments.

CUPID 코드를 이용한 CANDU 감속재의 열-유동 예비해석

박상기(S.G. Park),이재룡(J.R. Lee),윤한영(H.Y. Yoon),정재준(J.J. Jeong) 한국전산유체공학회 2012 한국전산유체공학회 학술대회논문집 Vol.2012 No.11

The objectives of this study are to predict CANDU moderator temperatures by using the CUPID code and to validate the CUPID against STERN 2D experimental data. In this study, both single and two phase flows in the Calandria vessel were calculated by using the CUPID code. KAERI has been developing the CUPID Code. It adopts three-dimensional, transient, two-phase and three-field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two-fluid model. The CUPID code validated using the single-phase flow experimental data that were performed in the STERN Lab.

CUPID 코드를 활용한 2X2 봉다발 부수로 유동 해석

이재룡(J.R. Lee),박익규(I.K. Park),김정우(J. Kim) 한국전산유체공학회 2016 한국전산유체공학회지 Vol.21 No.4

The CUPID code is a transient, three-dimensional, two-fluid, thermal-hydraulic code designed for a component-scale analysis of nuclear reactor components. The primary objective of this study is to assess the applicability of CUPID to single-phase turbulent flow analyses of 2×2 rod bundle subchannel. The bulk velocity at the inlet varies from 1.0 m/s up to 2.0 m/s which is equivalent to the fully turbulent flow with the range of Re=12,500 to 25,000. Adiabatic single-phase flow is assumed. The velocity profile at the exit region is quantitatively compared with both experimental measurement and commercial CFD tool. Three different boundary conditions are simulated and quantitatively compared each other. The calculation results of CUPID code shows a good agreement with the experimental data. It is concluded that the CUPID code has capability to reproduce the turbulent flow behavior for the 2x2 rod bundle geometry.

CUPID 코드를 이용한 연구로 사이펀 차단 현상 해석

박종필(J.P. Park),박익규(I.K. Park) 한국전산유체공학회 2019 한국전산유체공학회지 Vol.24 No.3

A pipe rupture of a primary cooling system in a open-pool type research reactor could lead to drainage of the coolant in a reactor pool though this pipe by siphon effect. As a consequence, the core could be exposed to the air. Therefore, this type of research reactor is equipped with a siphon breaking system to maintain reactor pool water level above the required height during a loss of coolant accident. In this study, numerical simulations are performed to assess the effect of interfacial drag on siphon break phenomena using CUPID code. The CUPID results indicate that siphonage break flow rate and initiation of siphon break are significantly affected by the interfacial drag. The calculated break flow rate and final water level in the pool using CUPID code with relevant interfacial drag are a good agreement with experimental data.

CUPID 코드를 활용한 ROCOM 붕소 혼합 문제 해석

조윤재,윤한영 한국전산유체공학회 2018 한국전산유체공학회지 Vol.23 No.4

The concentration of boron in the coolant of reactor cooling system in nuclear power plants(NPPs) plays an important role in controlling the reactivity according to fuel burn-up. Thus, an accurate prediction of the boron concentration after an unexpected de-borated water injection is required for the safety of NPPs. Korea Atomic Energy Research Institute(KAERI) participates in the ROCOM boron dilution benchmark in a framework of IAEA coordinated research projects. Through the benchmark, the capability of the CUPID code to predict the boron mixing behavior was validated. Two-equations RANS turbulence models were tested such as the standard k-ε model, RNG k-ε model, and SST k-ω model. The calculation results showed that the standard k-ε model reasonably predicted the boron mixing behavior when appropriate y+ values were verified even though the k-ε model required less number of cells comparing to the SST k-ω model because of larger y+ values.

CUPID코드를 활용한 2×2 봉다발 부수로 내부 단상 및 2상 난류유동 해석

이종혁,박익규,이재룡,김정우 한국전산유체공학회 2017 한국전산유체공학회지 Vol.22 No.4

It is important to understand the flow characteristic in a subchannel of nuclear rod bundle in the design process of a nuclear reactor considering the safety margin. In this study, the CUPID code was used to simulate the flow phenomena in the 2x2 rod bundle. The simulation of 2x2 rod bundle test under single- and two-phase flow conditions were conducted to validate the turbulence model. The liquid velocities at 0.45 m along the flow path with 0 and 45 degree of azimuthal angle were compared with experimental data. The CUPID results were a good agreement with experimental data, though the ones for 0 degree case has a little discrepancy. For the future work, the turbulent flow will be investigated by applying different turbulence models to improve the prediction of velocity near wall.

기기해석용 코드 CUPID를 이용한 코어캐쳐 실험장치의 전산성능예측

이동훈(D.H. Lee),박익규(I.K Park),윤한영(H.Y. Yoon),정재준(J.J. Jeong) 한국전산유체공학회 2012 한국전산유체공학회 학술대회논문집 Vol.2012 No.11

The analysis of a core catcher test facility was carried out using the CUPID, which is a three-dimensional thermal hydraulic code having been developed by KAERL The core catcher is considered as a promising engineered system to stabilize the molten corium emitted from the reactor vessel following core melt and rupture of reactor vessel during a postulated severe accident. Conceptually, the core catcher consists of a carbon steel body, sacrificial material, protection material, and engineered cooling channel. The cooling capacity of the engineered cooling channel should be guaranteed to remove the decay heat of the molten corium. The flow in the core catcher is a combined problem of a two-phase flow in the engineered cooling channel and a single phase natural circulation in the whole core catcher system. The computational analysis using the CUPID code showed that it can appropriately simulate the multi-dimensional boiling phenomena under a low pressure and low flow rate condition.