• Nuclear Engineering and Technology
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• 제27권6호
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• pp.859-869
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• 1995

본 연구는 가압경수로의 부분충수 운전중 잔열제거기능 상실사고 해석시 신뢰성을 확보하기 위해 RELAP5/MOD3.1 코드로 관련 대형 실험을 모의 계산하여, 사고시 예상되는 주요 물리적 현상의 파악과 코드의 예측능력을 평가하는 것이다. 대상 실험으로 선택된 BETHSY Test 6.9a는 이 사고중 증기발생기가 작동하지 않고, 가압기 Manway를 개방한 상태 (Configuration)를 모의한 실험이다. 이 연구 결과는 실제 원전 사고시 예상되는 중요 현상 뿐 아니라, 이에 영향을 미치는 민감한 인자를 파악하여 사고 해석결과의 유효성을 판단하는 데 상당히 기여할 것으로 기대한다. 연구결과 RELAP5/MOD3.1 코드는 대체적으로 계통의 과도기 거동은 타당하게 예측하고 있지만, 모의계산에서 Time-Step이 아주 짧아 막대한 시간이 소요된다는 문제점이 발견되었다. 그 외에도 노심팽창수위 (swollen level)를 과대평가하여 가압기의 수위 및 계통의 압력을 높게 계산하였다. 이로 인해 가압기를 통한 방출량도 과대계산하여 노심노출을 약 400초 빨리 예측하였다. 실험과 코드 예측결과를 종합할 때 가압기 Manway 만의 개방으로는 계통압력이 상승하고, 중력주입냉각수로는 노심수위 회복에 불충분하며, 결국 강제주입에 의해서 노심수위가 회복될 수 있음이 입증되었다.

• 한국안전학회지
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• 제28권3호
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.419-433
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• 2016

One of the important severe accident management measures in the Light Water Reactors is water injection to the reactor core. The related phenomena are investigated by performing experiments and computer simulations. One of the most widely known is the QUENCH test-program. A number of analyses on QUENCH tests have also been performed by different computer codes for code validation and improvements. Unfortunately, any deterministic computer simulation is not free from the uncertainties. To receive the realistic calculation results, the best estimate computer codes should be used for the calculation with combination of uncertainty and sensitivity analysis of calculation results. In this article, the QUENCH-03 and QUENCH-06 experiments are modelled using ASTEC and RELAP/SCDAPSIM codes. For the uncertainty and sensitivity analysis, SUSA3.5 and SUNSET tools were used. The article demonstrates that applying the best estimate approach, it is possible to develop basic QUENCH input deck and to develop the two sets of input parameters, covering maximal and minimal ranges of uncertainties. These allow simulating different (but with the same nature) tests, receiving calculation results with the evaluated range of uncertainties.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1306-1313
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• 2018

The randomness and incipient nature of certain faults in reactor systems warrant a robust and dynamic detection mechanism. Existing models and methods for fault diagnosis using different mathematical/statistical inferences lack incipient and novel faults detection capability. To this end, we propose a fault diagnosis method that utilizes the flexibility of data-driven Support Vector Machine (SVM) for component-level fault diagnosis. The technique integrates separately-built, separately-trained, specialized SVM modules capable of component-level fault diagnosis into a coherent intelligent system, with each SVM module monitoring sub-units of the reactor coolant system. To evaluate the model, marginal faults selected from the failure mode and effect analysis (FMEA) are simulated in the steam generator and pressure boundary of the Chinese CNP300 PWR (Qinshan I NPP) reactor coolant system, using a best-estimate thermal-hydraulic code, RELAP5/SCDAP Mod4.0. Multiclass SVM model is trained with component level parameters that represent the steady state and selected faults in the components. For optimization purposes, we considered and compared the performances of different multiclass models in MATLAB, using different coding matrices, as well as different kernel functions on the representative data derived from the simulation of Qinshan I NPP. An optimum predictive model - the Error Correcting Output Code (ECOC) with TenaryComplete coding matrix - was obtained from experiments, and utilized to diagnose the incipient faults. Some of the important diagnostic results and heuristic model evaluation methods are presented in this paper.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.627-636
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• 2007

The present work is aimed at further discussing the effectiveness of dimensionless parameters recently proposed for the analysis of flow stability in heated channels with supercritical fluids. In this purpose, after presenting the main motivations for the introduction of these parameters in place of previously proposed ones, additional information on the theoretical bases and on the consequences of this development is provided. Stability maps, generated by an in-house program adapted from a previous application to boiling channels, are also shown for different combinations of the operating parameters. The maps are obtained as contour plots of an amplification parameter obtained from numerical discretization and subsequent linearization of governing equations; as such, they provide a quantitatively clear perspective of the effect of different boundary conditions on the stability of heated channels with supercritical fluids. In order to assess the validity of the assumptions at the basis of the in-house model, supporting calculations have been performed making use of the RELAP5/MOD3.3 computer code, detecting the values of the dimensionless parameters at the threshold for the occurrence of instability for a heated channel representative of SCWR proposed core configurations. The obtained results show reasonable agreement with the maps, supporting the applicability of the proposed scaling parameters for describing the dynamic behaviour of heated channels with supercritical fluids.

• Nuclear Engineering and Technology
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• 제18권2호
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• pp.97-106
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• 1986

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.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1791-1801
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• 2023

The multi-physics coupled methodologies that have been widely used to analyze the complex process occurring in nuclear reactors have also been used to the R&D of numerical reactors. The advancement in the field of computer technology has helped in the development of these methodologies. Herein, we report the integration of ADPRES code and RELAP5 code into the SALOME-ICoCo framework to form a multi-physics coupling platform. The platform exploits the supervisor architecture, serial mode, mesh one-to-one correspondence and explicit coupling methods during analysis, and the uncertainty analysis tool URANIE was used. The correctness of the platform was verified through the NEACRP-L-335 benchmark. The results obtained were in accordance with the reference values. The platform could be used to accurately determine the power peak. In addition, design margins could be gained post uncertainty analysis. The initial power, inlet coolant temperature and the mass flow of assembly property significantly influence reactor safety during the rod ejections accident (REA).

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1955-1962
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• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.

• Nuclear Engineering and Technology
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• 제25권1호
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• pp.37-50
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• 1993

완전급수상실사고 및 복구가정의 사고전개와 열수력학적 거동을 평가하기 위해 RELAP5/MOD3 계산을 수행하고 LOFT L9-l/L3-3 실험 결과와 비교하였다. 또한 본 사고의 주요 열수력 현상에 대한 코드의 예측도를 평가하였다. 본 연구의 결과로서 가압기 수위가 만수위에 도달할 때까지 살수를 이용한 압력 제어, 가압기 압력방출 밸브를 통한 가압방지, 증기발생기 재충수에 의한 이차측 열제거 능력의 재확보, 지속적인 자연순환에 의한 효과적인 일차계통의 냉각등이 이루어 질 수 있고 이 과정중 노심노출은 나타나지 않음이 밝혀졌다. 또한 현재의 현상-중심 비상운전절차서 특히 과압방지성능 및 증기발생기 회복절차 등의 유효성을 입증하기 위해서는 발전소 고유한 평가가 필요함이 밝혀졌다.

• 시스템엔지니어링학술지
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• 제18권2호
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• pp.94-107
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• 2022

Accidents prevention and mitigation is the highest priority of nuclear power plant (NPP) operation, particularly in the aftermath of the Fukushima Daiichi accident, which has reignited public anxieties and skepticism regarding nuclear energy usage. To deal with accident scenarios more effectively, operators must have ample and precise information about key safety parameters as well as their future trajectories. This work investigates the potential of machine learning in forecasting NPP response in real-time to provide an additional validation method and help reduce human error, especially in accident situations where operators are under a lot of stress. First, a base-case SGTR simulation is carried out by the best-estimate code RELAP5/MOD3.4 to confirm the validity of the model against results reported in the APR1400 Design Control Document (DCD). Then, uncertainty quantification is performed by coupling RELAP5/MOD3.4 and the statistical tool DAKOTA to generate a large enough dataset for the construction and training of neural-based machine learning (ML) models, namely LSTM, GRU, and hybrid CNN-LSTM. Finally, the accuracy and reliability of these models in forecasting system response are tested by their performance on fresh data. To facilitate and oversee the process of developing the ML models, a Systems Engineering (SE) methodology is used to ensure that the work is consistently in line with the originating mission statement and that the findings obtained at each subsequent phase are valid.