• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.596-609
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• 2002

Korea Electric Power Research Institute has developed the in-house safety analysis methodologies for non-LOCA(Loss Of Coolant Accident) events based on codes and methodologies of vendors and Electric Power Research Institute . According to the new methodologies, analyses of system responses and calculation of DNBR(Departure from Nucleate Boiling Ratio) during the transient have been carried out with RETRAN code and a sub-channel analysis code, respectively. However, it takes too much time to calculate DNBR for each case using the two codes to search for the limiting case from sensitivity study. To simplify the search for the limiting case, accordingly, RETRAN code has been modified to roughly calculate DNBR using hot channel modeling. The W-3 correlation is already included in RETRAN as one of the auxiliary DNBR models. However, WRB-1 and WRB-2 correlations required to analyze some Westinghouse type fuels are not considered in RETRAN DNBR models. In this paper, the RETRAN DNBR models using the correlations have been developed and the partial and complete loss of forced reactor coolant flow events have been analyzed for Yonggwang units 1 and 2 with the new methodologies to validate the models. The results of the analyses have been compared with those mentioned in the chapter 15 of the Final Safety Analysis Report.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.389-395
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• 1995

기존의 천이노심 DNBR 벌점 평가 방법을 개선하여 불확실도를 줄이고 신뢰도를 향상시키며, 적용범위를 확대함으로써 보다 실제적인 DNBR 벌점 평가 방법을 제시하고자 하였다. 이를 위하여 영광 1호기 JDFA-V5H 의 천이노심을 대상으로 하는 일련의 분석이 수행되었다. 먼저 균일노심과 천이노심을 모형화 한 기준 제어군에서의 상대적인 물성치의 변화와 축방향에서의 DNBR 거동을 분석하였고 이에 따른 최소 DNBR 의 상대적 차이로부터 최대 럴점 조건 및 벌점이 적용되는 집합체를 선정하였다. 변수 민감도 분석 결과, 최대 벌점 조건은 과출력 (120% 출력), 고압 (2420 psia) 그리고 상부노심에서 상대출력이 많은 축방향 출력 분포를 갖는 조건이 선정되었고 천이노심 벌점은 V5H에만 부과된다. 천이노심 DNBR 벌점은 배열 민감도 분석을 통하여 노심내 V5H 분율의 함수로 표현됨을 알 수 있었으며, 기존의 보수적인 방법론에 비해 최소 3% 이상의 천이노심 벌점이 감소되는것으로 나타나 추가적인 여유도의 확보로 인한 설계의 탄력성을 기대할 수 있다. 이 결과는 IFM이 존재하는 원전연료 집합체 상부에 대하여 노심의 V5H 분율이 0.02 부터 1.0 까지의 정상 및 과도상태 노심에 대하여 적용할 수 있다.

• Nuclear Engineering and Technology
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• v.30 no.6
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• pp.617-627
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• 1998

A technically more direct Statistical Combination of Uncertainties (SCU) method, extended SCU (XSCU), was developed to statistically combine the uncertainties associated with the DNBR alarm setpoint and the DNBR trip setpoint of digital nuclear power plants. The Modified SCU (MSCU) method is currently used as the USNRC approved design method to perform the same function. In this study, the MSCU and XSCU methods were compared in terms of the total uncertainties, and the thermal margins to the DNBR alarm and trip setpoints. The MSCU method resulted in small total uncertainties due to large negative biases which are unphysical. The XSCU method gives virtually unbiased total uncertainties which are physically meaningful in order to represent the actual magnitude of the total uncertainties associated with the DNBR alarm and trip setpoints. But the thermal margins to the DNBR alarm and trip setpoints by the MSCU method agree with those by the XSCU method within allowable statistical Variations.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.133-133
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.160.1-160
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• 2001

• 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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• 2001.05a
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• pp.148.1-148
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• 2001

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.236-247
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• 1983

Analyzed are the the DNBR(Departure from Nucleate Boiling Ratio) sensitivities to variations in various PWR operating parameters utilizing the Korea Nuclear Unit 1(KNU-1) design and operating data. Studied parameters in the analysis are core power level, system pressure, core inlet flow rate, core inlet temperature, enthalpy rise hot channel factor, and axial power peaking factor and axial offset. The calculations are performed using the steady state and transient thermal-hydraulics computer program, COBRA-IV-K, which is the revised version of COBRA-IV-i that has been adapted, partially modified and verified at KAERI. A reference case is established based on the design and operating condition of the KNU-1 reactor core, and this provides a basis for the subsequent sensitivity analysis. From the calculation results it is concluded that the most sensitive parameter in the DNBR thermal design is the coolant core inlet temperature while the axial power peaking factor is the least sensitive.