The Prediction of Void Fraction in the Subcooled Boiling Region

Goon Cherl Park Korean Nuclear Society 1984 Nuclear Engineering and Technology Vol.16 No.4

축방향 비균일 열유량을 가지는 서브쿨드 비등영역에서 첨단적인 mechanistic 모델을 사용하여 기포계수를 정확히 계산하고저 한다. 이 모델에서는 Lahey/Ohama model에서 구한 기포계수에 의존하는 drift-flux계수를 사용하였고 또한 실제 실험치와 비교하여 질량유량에 의존하는 응축계수의 상관식을 구하여 사용하였다. 이 모델은 고 정확도를 증명하기 위해 잘 알려진 실험치들과 비교되었고 최종적으로 고리 1호기 1주기의 고온 연료집합체 기포계수를 계산하므로서 profilefit 모델과 비교되었다. 이러한 계산결과는 실험치와 잘 일치하고 있으며 profile-fit model은 서브쿨드 비등영역에서 기포 계수를 낮게 계산하고 있음이 밝혀졌다. A state-of-the-art mechanistic model has been developed to accurately predict the void fraction in the subcooled boiling region having axial nonuniform heat flux. In this study, the void-dependent drift-flux parameters of the Lahey/Ohkawa model were introduced and the mass flux-dependent condensation coefficient were determined by fitting with the experimental data. This model was tested against several experimental data sets to verify its accuracy. Finally the comparison between the predicted void fraction profiles with this model and the profile-fit model for the hot assembly of Kori-Unit 1, Cycle 1 has been performed. It is conclusive that the results show the good agreement between the measured and predicted void fractions, and the profile-fit model has been found to underestimate the void fraction in the subcooled boiling region.

원자력 열수력 실험 연구의 현황과 미래

박군철(Goon-Cherl Park),전지한(Ji-Han Chun) 대한기계학회 2009 大韓機械學會論文集B Vol.33 No.9

This paper introduces the current nuclear experimental research activities in KAERI, the unique nuclear research institute in Korea, and the universities in Korea to solve and assess the issues which have been faced in the design of new reactors such as APR1400, SMART, GEN-IV reactors as well as fusion reactor. Also the experimental evaluations of safety for operating nuclear plants have been presented. The nuclear thermalhydraulic experiments performed in such organizations are classified the fundamental test, the separated effect test, and the integral effect test with ATLAS and SNUF. Introduction is deployed according to institutes. Finally, the future works and the direction of research voyage in the nuclear thermal-hydraulic field were suggested.

ASSESSMENT OF A NEW DESIGN FOR A REACTOR CAVITY COOLING SYSTEM IN A VERY HIGH TEMPERATURE GAS-COOLED REACTOR

PARK GOON-CHERL,CHO YUN-JE,CHO HYOUNGKYU Korean Nuclear Society 2006 Nuclear Engineering and Technology Vol.38 No.1

Presently, the VHTGR (Very High Temperature Gas-cooled Reactor) is considered the most attractive candidate for a GEN-IV reactor to produce hydrogen, which will be a key resource for future energy production. A new concept for a reactor cavity cooling system (RCCS), a critical safety feature in the VHTGR, is proposed in the present study. The proposed RCCS consists of passive water pool and active air cooling systems. These are employed to overcome the poor cooling capability of the air-cooled RCCS and the complex cavity structures of the water-cooled RCCS. In order to estimate the licensibility of the proposed design, its performance and integrity were tested experimentally with a reduced-scale mock-up facility, as well as with a separate-effect test facility (SET) for the 1/4 water pool of the RCCS-SNU to examine the heat transfer and pressure drop and code capability. This paper presents the test results for SET and validation of MARS-GCR, a system code for the safety analysis of a HTGR. In addition, CFX5.7, a computational fluid dynamics code, was also used for the code-to-code benchmark of MARS-GCR. From the present experimental and numerical studies, the efficacy of MARS-GCR in application to determining the optimal design of complicated systems such as a RCCS and evaluation of their feasibility has been validated.

A Study on the Coolant Mixing Phenomena in the Reactor Lower Plenum

Yong Seog Choi 한국원자력학회 1997 Nuclear Engineering and Technology Vol.29 No.3

원전의 부분충수운전에 대한 동적 신뢰도평가

박군철,제무성 한국산업안전학회 1996 한국안전학회지 Vol.11 No.2

This paper presents a new approach for assessing the dynamic reliability in a complex system such as a nuclear power plant. The method is applied to a dynamic analysis of the potential accident sequences which may occur during mid-loop operation. Mid-loop operation is defined as an operation to make RCS water level below the top of the flow area of the hot legs at the junction with the reactor vessel for repairs and maintenance of steam generators and reactor coolant pumps for a specific time. The idea behind this approach consists of both the use of the concept of the performance achievement/requirement correlation and of a dynamic event tree generation method. The assessment of the system reliability depends on the determination of both the required performance distribution and the achieved performance distribution. The quantified correlation between requirement and achievement represents a comparison between two competing variables. It is demonstrated that this method is easily applicable and flexible in that it can be applied to any kind of dynamic reliability problem.

Generalized Nyquist Criterion for the Stability of Xenon Oscillation

Park, You-Cho,Park, Goon-Cherl,Chung, Chang-Hyun,Park, Chong-Kyun Korean Nuclear Society 1990 Nuclear Engineering and Technology Vol.22 No.4

The Xenon spatial oscillation may give rise to operational difficulties in a nuclear power plant. In this study, in order to investigate the Xenon instability for a PWR, the frequency-domain technique is adopted by using Generalized Nyquist Criterion, which is more general and suitable for the multi-input/multi-output system. Also linearized modal fluxes are obtained by a modal expansion. This model has been implemented to test the axial Xenon stability of YGN-1 unit against the changes in plant operating parameters ; power level, control rod position, and core average burnup. The results show that the increase of power level and the deeper insertion of control rod have the destabilizing effect, and that the burnup progress makes the core less stable. Also the results show that the overestimation due to modal interaction was found not to be significant.

LOCA Analysis and Development of a Simple Computer Code for Refill-Phase Analysis

Ree, Hee-Do,Park, Goon-Cherl,Kim, Hyo-Jung,Kim, Jin-Soo Korean Nuclear Society 1986 Nuclear Engineering and Technology Vol.18 No.3

The loss of coolant accident based on a double-ended cold leg break is analyzed with the discharge coefficient (Ca) of 0.4. This analysis covers the whole transient period from the start of depressurization to the complete refilling of the core by using RELAP4/MOD6-EM and RELAP4/ MOD6-HOT CHANNEL for the system thermal-hydraulics and the fuel performance during the blowdown phase respectively, and RELAP4/MOD6-FLOOD and TOODEE2 during the reflood phase. A simple analytical method has been developed to account for the lower plenum filling by approximating steam-water countercurrent flows and superheated wall effects at the downcomer during the refill period. Based on the informations. at the time of EOB (end-of-bypass), the refill duration time and the initial reflooding temperature were estimated and compared with the results from the RELAP4/MOD6, resulting in a good agreement. In addition, some parametric studies on the EOB were performed. The form loss coefficient between upper head and upper downcomer was found to be sensitive to the occurrence of the spurious EOB. Appropriate form loss coefficients should be taken into account to avoid the flow oscillations at the downcomer. The analyses with the six and three volume core nodalizations, respectively, show much similar trends in the system thermal-hydraulic performance, but the former case is recommended to obtain good results.

Multiphase Flow Modeling of Molten Material-Vapor-Liquid Mixtures in Thermal Nonequilibrium

Park, Ik-Kyu,Park, Goon-Cherl,Bang, Kwang-Hyun The Korean Society of Mechanical Engineers 2000 JOURNAL OF MECHANICAL SCIENCE AND TECHNOLOGY Vol.14 No.5

This paper presents a numerical model of multi phase flow of the mixtures of molten material-liquid-vapor, particularly in thermal nonequilibrium. It is a two-dimensional, transient, three-fluid model in Eulerian coordinates. The equations are solved numerically using the finite difference method that implicitly couples the rates of phase changes, momentum, and energy exchange to determine the pressure, density, and velocity fields. To examine the model's ability to predict an experimental data, calculations have been performed for tests of pouring hot particles and molten material into a water pool. The predictions show good agreement with the experimental data. It appears, however, that the interfacial heat transfer and breakup of molten material need improved models that can be applied to such high temperature, high pressure, multi phase flow conditions.

조밀화 집합체로 중간저장하는 경우 원자력 발전소 9, 10호기의 사용 후 핵연료 저장조의 임계분석

Jae, Moo-Sung,Park, Goon-Cherl,Chung, Chang-Hyun,Jang, Jong-Hwa Korean Nuclear Society 1988 Nuclear Engineering and Technology Vol.20 No.1

Since the lack of the spent fuel storage capcity has been expected for all Korean nuclear power plants in the mid-1990s, the maximum density rack (MDR) with consolidated fuels can be proposed to overcome the shortage of the storage capacity in KNU 9 & 10 which have most limited capacities. To ensure the safety when the alternatives are applied in the KNU 9 & 10, the multiplication factor are calculated with varying the rack pitch and the thickness of consolidated storage box by the AMPX-KENO IV codes. The computing system is verified by the benchmark calculation with criticality experiments for arrays of consolidated fuel modules, which was reported by B & W in 1981. Also an abnormal condition, i.e. malposition accident, is simulated. The results indicate that the KNU 9 & 10 storage pools with consolidated fuel are safe in the view of the criticality. Thus the storage capacity can be expanded from 9/3 cores into 27/3 cores even with considering equipments and cooling spaces.

The Analysis of the Cooling Capability for Consolidated Spent Fuel Pool in Uljin 1&2

Jo, Seong-Youn,Park, Goon-Cherl,Chung, Chang-Hyun 서울대학교 공과대학 1988 공대연구보고 Vol.20 No.2