• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.41-48
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• 2018

This paper describes the systems engineering development process for the Departure from Nucleate Boiling Ratio (DNBR) algorithm using FPGA. Current Core Protection Calculator System (CPCS) requirement and DNBR logic are analyzed in the reverse engineering phase and the new FPGA based DNBR algorithm is designed in the re-engineering phase. FPGA based DNBR algorithm is developed by VHSIC Hardware Description Language (VHDL) in the implementation phase and VHDL DNBR software is verified in the software Verification & Validation phase. Test cases are developed to perform the software module test for VHDL software modules. The APR 1400 simulator is used to collect the inputs data in 100%, 75%, and 50% reactor power condition. Test input signals are injected to the software modules following test case tables and output signals are compared with the expected test value. Minimum DNBR value from developed DNBR algorithm is validated by KEPCO E&C CPCS development facility. This paper summarizes the process to develop the FPGA-based DNBR calculation algorithm using systems engineering approach.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.503-517
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• 1995

A mechanistic model for forced convective transition boiling has been developed to predict transition boiling heat flux realistically. This model is based on a postulated multi­stage boiling process occurring during the passage time of an elongated vapor blanket specified at a critical heat flux condition. Between the departure from nucleate boiling (DNB) and the departure from film boiling (DFB) points, the boiling heat transfer is established through three boiling stages, namely, the macrolayer evaporation and dryout governed by nucleate boiling in a thin liquid film and the unstable film boiling. The total heat transfer rate during the transition boiling is the sum of the heat transfer rates after the DNB weighted by the time fractions of each stage, which are defined as the ratio of each stage duration to the vapor blanket passage time. The model predictions are compared with some available experimental transition boiling data. From these comparisons, it can be seen that the transition boiling heat fluxes including the maximum heat flux and the minimum film boiling heat flux are nil predicted at low qualities/high pressures near 10 bar.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.47-58
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• 2020

Most of the reactor trips in Korean NPPs related to core protection systems were caused not because of proximity of boiling crisis and, consequently, a damage in the core, but due to particular miscalculations or component failures related to the core protection system. The most common core protection system applied in Korean NPPs is the Core Protection Calculator System (CPCS), which is installed in OPR1000 and APR1400 plants. It generates a trip signal to scram the reactor in case of low Departure from Nucleate Boiling Ratio (DNBR) or high Local Power Density (LPD). However, is a reactor trip necessary to protect the core? Or could a fast power reduction be enough to recover the DNBR/LPD without a scram? In order to analyze the online calculation of DNBR/LPD, and the use of fast power reduction as trip avoidance methodology, a concept of CPCS with fast power reduction function was developed in Matlab® Simulink using systems engineering approach. The system was validated with maximum of 0.2% deviation from the reference and the dynamic deviation was maximum of 12.65% for DNBR and 6.72% for LPD during a transient of 16,000 seconds.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.508-513
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• 1998

가압경수로에서 열여유도를 평가하는 당대의 접근 방범이 고찰되었다. 통상적으로 열여유도 평가는 국부 열속이 임계열속(CHF: Critical Heat Flux)으로부터 떨어진 거리로부터 도출된 핵비등이탈률(DNBR: Departure from Nucleate Boiling Ratio) 개념을 사용하여 수행된다. 본 연구에서는 열여유도 정가에 대하여 제기되는 문제가 당대의 접근 방법에 이러한 영향을 주는 지 평가하고 이에 대한 향후 연구 방향을 살펴 보고자 하는 것이다. 혼합날게 그리드에 의해 야기되는 와류 효과가 크면 클수록 현재의 부수로 분석 방법의 신뢰성이 제고되며 상관식의 예측 성능을 평가할 때 데이터 분포가 중요하다는 착안점이 얻어졌다.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.24-32
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• 2018

System Complexity one of the most common cause failure of the projects, it leads to a lack of understanding about the functions of the system. Hence, the model is developed for communication and furthermore modeling help analysis, design, and understanding of the system. On the other hand, the text-based specification is useful and easy to develop but is difficult to visualize the physical composition, structure, and behaviour or data exchange of the system. Therefore, it is necessary to transform system description into a diagram which clearly depicts the behaviour of the system as well as the interaction between components. According to the International Atomic Energy Agency (IAEA) Safety Glossary, The safety system is a system important to safety, provided to ensure the safe shutdown of the reactor or the residual heat removal from the reactor core, or to limit the consequences of anticipated operational occurrences and design basis accidents. Core Protection Calculator System (CPCS) in Advanced Power Reactor 1400 (APR 1400) Nuclear Power Plant is a safety critical system. CPCS was developed using systems engineering method focusing on Departure from Nuclear Boiling Ratio (DNBR) calculation. Due to the complexity of the system, many diagrams are needed to minimize the risk of ambiguities and lack of understanding. Using Model-Based Systems Engineering (MBSE) software for modeling the DNBR algorithm were used. These diagrams then serve as the baseline of the reverse engineering process and speeding up the development process. In addition, the use of MBSE ensures that any additional information obtained from auxiliary sources can then be input into the system model, ensuring data consistency.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.544-549
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• 1995

An analysis for Steam Line Break (SLB) events which result in a return-to-power conditions after reactor trip was performed for a postulated Yonggwang Nuclear Power Plant Unit 3 cycle 8. Analysis methodology for post-trip return-to-power SLB is quite different from that of a no return-to-power SLB and is more complicated. Therefore, it is necessary to develop an methodology to analyze the response of the NSSS parameter and the fuel performance for the post-trip return-to-power SLB events. In this analysis, the cases with and without offsite power were simulated by crediting 3-D reactivity feedback effect due to local heatup around stuck CEA and compared with the cases without 3-D reactivity feedback with respect to fuel performance, departure from nucleate boiling ratio (DNBR) and linear heat generation rate (LHGR).

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.316-321
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• 1997

The mixing vane effect on the Critical Heat Flux (CHF) is discussed with focus on the vortex now effect. In the subchannel approach, this effect is not quantified by the calculation model, but directly taken into account by the CHF correlation itself through data analysis. The vortex now effect is identified the two Westinghouse correlations, and then the CHF margin issue given rise to by the Vantage-5H design change is evaluated and discussed. It is noted that deficiency about CHF dependency on the vortex flow effect could induce an error in the Departure from Nucleate Boiling Ratio (DNBR) sensitivity Calculation.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.237-246
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• 1994

A high flow critical heat flux (CHF) correlation, based on the single-pin CHF experimental data for finned and unfinned heated rods, was developed for the thermal-hydraulic design and safety analysis of the Korea Multi-purpose Research Reactor (KMRR) core. The correlation consists of dimensionless parameters such as Reynolds number, thermodynamic equilibrium quality, liquid-to-vapor density ratio, and hydraulic equivalent diameter ratio. The fin effect was taken into account in the correlation by a finned-to-unfinned heated perimeter ratio. The effects of a cold wall and non-uniform axial power distribution ore discussed to verify the applicability of the single-pin based correlation to the KMRR fuel bundle. The correlation limit departure from nucleate boiling ratio (DNBR) was determined as 1.44 from the statistical analysis of the CHF data.

• Journal of Energy Engineering
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• v.24 no.1
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• pp.149-163
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• 2015

In this study, we conducted study on the confirmation of thermal-hydraulic safety for Mast assembly with Computational Fluid Dynamics(CFD) analysis. Before performing the natural convection analysis for the Mast assembly by using CFD code, we validated the CFD code against two benchmark natural convection data for the evaluation of turbulence models and confirmation of its applicability to the natural convection flow. From the first benchmark test which was performed by Betts et al. in the simple rectangular channel, we selected standard k-omega turbulence model for natural convection. And then, calculation performance of CFD code was also investigated in the sub-channel of rod bundle by comparing with PNL(Pacific Northwest Laboratory) experimental data and prediction results by MATRA and Fluent 12.0 which were performed by Kwon et al.. Finally, we performed main natural convection analysis for fuel assembly inside the Mast assembly by using validated turbulence model. From the calculation, we observed stable natural circulation flow between the mast assembly and pool side and evaluated the thermal-hydraulic safety by calculating the departure from nucleate boiling ratio.