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고리 1호기 외부 전원 상실사고에 의한 RELAP5/MOD2코드 모델 평가
Chung, Bub-Dong,Kim, Hho-Jung,Lee, Young-Jin Korean Nuclear Society 1990 Nuclear Engineering and Technology Vol.22 No.1
The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip.
B$\Phi$rrensen Model Computation for Neutronic Benchmark Problems
Bub Dong Chung,Chang Hyo Kim,Chang Hyun Chung Korean Nuclear Society 1981 Nuclear Engineering and Technology Vol.13 No.2
B$\Phi$rrensen은 3차원적 노심분석에 긴 시간이 요구되는 유한차분법의 대안으로서 경수형 원자로의 전체출력분포를 계산하는데 한가지의 coarse mesh 방법을 제안하였다. 이 방법은 계산시간이 매우 짧은 것으로 알려져 있지만. 그 계산의 정확도에 대한 것은 아직까지 알려져 있지 않다. 본 논문에서는 B$\Phi$rrensen의 방법을 IAEA benchmark 문제와 RIS$\Phi$ benchmark 문제에 적용시켜서 계산시간과 정확도에 대해서 고찰하였다. 두 문제에서 노심-반사체와의 경계조건과 B$\Phi$rrensen의 모델상수들의 개선으로 출력분포 계산의 정확도를 상당히 높힐 수 있었다. B$\Phi$rrensen proposed a coarse mesh, three-dimensional one-and-half group diffusion scheme for computing the gross power distribution in light water reactors as an alternative to the conventional fine mesh finite difference approach in dealing with three dimensional problems, which require a prohibitively long computing time. The method reported takes extremely small execution time. However, its computational accuracy has not been investigated yet. The B$\Phi$rrensen method is revised in this work and both efficiency and accuracy are examined by applying it to IAEA benchmark problem and RIS$\Phi$ benchmark problem. It is found that two modifications on core-reflector boundary conditions and B$\Phi$rrensen's model constants may improve computational accuracy of power distribution calculation.
Kim, Hho-Jung,Chung, Bub-Dong,Lee, Young-Jin,Kim, Jin-Soo Korean Nuclear Society 1986 Nuclear Engineering and Technology Vol.18 No.2
1981년 6일 9일 원자력 1호기에서 발생한 77.5% 출력상태에서의 외부전원상실사고를 열, 수력학적최적계산용 코드인 RELAP5/MODl/NSC를 사용하여 모의하였으며 해석결과는 발전소 실측자료와 잘 일치하였다. 원자로 냉각재펌프의 트립에 따른 flow coastdown후에 hot-cold leg온도차에 의하여 자연순환 유동이 형성됨이 확인되었으며 실측자료와 잘 일치하여 이와 관련된 전산코드의 열수력학 적모델의 타당성을 입증할 수 있었다. 또한 위의 사고전개가 정상운전상태인 전출력(100%)에서 재발하였을 경우를 가정하여 해석하였다. 이러한 해석을 통하여 보조급수의 공급과 더불어 증기발생기 PORV의 적절한 작동으로 원자력 1호기 노심잔열을 제거하여 안전성에 문제점을 야기하지 않음을 입증하였다. 최적 계산방법에 의한 사고해석에서는 turbine stop valve 작동시간, 증기 발생기 PORV 설정치 등 non-safety 관련요소들의 특성에 대한 정화한 모의가 필수적이다.
Analysis of Loss of Normal Feedwater Transient Using RBLAP5/MOD1/NSC; KNU1 Plant Simulation
Hho Jung Kim,Bub Dong Chung,Young Jin Lee,Jin Soo Kim Korean Nuclear Society 1986 Nuclear Engineering and Technology Vol.18 No.1
1984년 11월 14일 원자력 1호기에서 발생된 주급수 상실사고에 대한 계통의 열수력학적인 거동을 모의·해석하고, 발전소 실측자료와의 비교를 통하여 사용된 전산코드의 신뢰도를 평가하였다. 모의된 열수력학적 변수들은 발전소 실측자료와 비교적 잘 일치하였으나 원자로 트립시에 증기발생기 증기유량과 주 냉각재 계통 평균온도에 있어서 약간의 차이를 보였다. 이는 원자로 트립시 깎은 시간에 급격한 노심 출력의 감소로 인하여 열·수력학적 변수들에 큰 변화를 야기하여 발전소 실측자료가 과도상태에서의 불학실성을 내포하기 때문으로 예측되었다. 해석에 사용된 전산코드는 RELAP5/MOD1/CY018로부터 불합리한 oscillation을 일으키는 interphase drag 및 wall heat transfer model의 수정을 통하여 개발된 RELAP5/MOD1/NSC이다. Simulation of the system thermal-hydraulic parameters was carried out following the KNUl (Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on November 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS (Reactor Coolant System) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018.
Ahn, Kwang-Il,Chung, Bub-Dong,Lee, John C. Korean Nuclear Society 2010 Nuclear Engineering and Technology Vol.42 No.2
As pointed out in the OECD BEMUSE Program, when a high computation time is taken to obtain the relevant output values of a complex physical model (or code), the number of statistical samples that must be evaluated through it is a critical factor for the sampling-based uncertainty analysis. Two alternative methods have been utilized to avoid the problem associated with the size of these statistical samples: one is based on Wilks' formula, which is based on simple random sampling, and the other is based on the conventional nonlinear regression approach. While both approaches provide a useful means for drawing conclusions on the resultant uncertainty with a limited number of code runs, there are also some unique corresponding limitations. For example, a conclusion based on the Wilks' formula can be highly affected by the sampled values themselves, while the conventional regression approach requires an a priori estimate on the functional forms of a regression model. The main objective of this paper is to assess the feasibility of the ACE-RSM approach as a complementary method to the Wilks' formula and the conventional regression-based uncertainty analysis. This feasibility was assessed through a practical application of the ACE-RSM approach to the LOFT L2-5 LBLOCA PCT uncertainty analysis, which was implemented as a part of the OECD BEMUSE Phase III program.
RELAP5/MOD2 코드에 의한 대형냉각재 상실사고 모사실험 L2-3의 열수력 현상 예측
Bang, Young-Seok,Chung, Bub-Dong,Kim, Hho-Jung Korean Nuclear Society 1991 Nuclear Engineering and Technology Vol.23 No.1
The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical How model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5 / MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena.
Uncertainty Quantification of RELAP5/MOD3/KAERI on Reflood Peak Cladding Temperature
Park, Chan-Eok,Chung, Bub-Dong,Lee, Young-Jin,Lee, Guy-Hyung,Lee, Sang-Yong Korean Nuclear Society 1994 Nuclear Engineering and Technology Vol.26 No.3
The predictability of KAERI version of RELAP5/MOD3 on reflood peak cladding temperature during large break loss-of-coolant accident is assessed against 18 test runs in FLECHT SEASET test data. The associated uncertainty is statistically quantified. The selected test runs include a gravity feed test and several forced feed tests with wide range of the parameters such as flooding rate, system pressure, initial clad temperature, rod bundle power. The results show that the code under-predicts the peak cladding temperature by 7.56 K on average. The upper limit of the associated uncertainty at 95% confidence level is evaluated to be about 99 K, It including the bias due to the under-prediction.
Bae, Sung-Won,Chung, Bub-Dong Korean Nuclear Society 2009 Nuclear Engineering and Technology Vol.41 No.10
A multi-dimensional component for the thermal-hydraulic system analysis code, MARS, was developed for a more realistic three-dimensional analysis of nuclear systems. A three-dimensional and two-fluid model for a two-phase flow in Cartesian and cylindrical coordinates was employed. The governing equations and physical constitutive relationships were extended from those of a one-dimensional version. The numerical solution method adopted a semi-implicit and finite-difference method based on a staggered-grid mesh and a donor-cell scheme. The relevant length scale was very coarse compared to commercial computational fluid dynamics tools. Thus a simple Prandtl's mixing length turbulence model was applied to interpret the turbulent induced momentum and energy diffusivity. Non drag interfacial forces were not considered as in the general nuclear system codes. Several conceptual cases with analytic solutions were chosen and analyzed to assess the fundamental terms. RPI air-water and UPTF 7 tests were simulated and compared to the experimental data. The simulation results for the RPI air-water two-phase flow experiment showed good agreement with the measured void fraction. The simulation results for the UPTF downcomer test 7 were compared to the experiment data and the results from other multi-dimensional system codes for the ECC delivery flow.