• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.189-196
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• 1988

This study investigates the effect of fuel pellet eccentricity on fuel rod thermal performance under the steady state condition. The governing equations in the fuel pellet and the cladding region are set up in 2-dimensional cylindrical coordinate (r, $\theta$) and are solved by finite element method. The angular-dependent heat transfer coefficient in the gap region is used in order to account for the asymmetry of gap width. Material propeties are used as a function of temperature and volumetric heat generation as a function of radial position. The results show the increase of maximum local heat flux at the cladding outer surface and the decrease of maximum and average fuel temperatures due to eccentricity. The former is expected to affect the uncertainties in the minimum DNBR calculation. The latter two are expected to reduce the possibility of fuel melting and the fuel stored energy. Also, the fuel pellet eccentricity introduces asymmetry in fuel pellet temperature and movement of the location of maximum fuel pellet temperature.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Nuclear Engineering and Technology
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• v.14 no.3
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• pp.122-137
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• 1982

Analyzed is the sensitivity of reactor transient behavior to various reactor parameters during the reactivity induced accidents (RIA) of the Kori Unit 1. Included in the analysis is a partial spectrum of RIAs with relatively fast transients such as uncontrolled rod cluster control assembly bank withdrawl from a subcritical or low power startup condition and rod ejection accidents. The analysis can be performed generally in three steps: calculation of an average core power change, hot spot heat transfer calculation and DNBR (departure from nucleate boiling ratio) calculation. The computer codes used for the analysis are either developed based on the codes relevent to it. These codes are evaluated to be highly reliable. An extensive sensitivity analysis is performed to study the effects of various reactor design and operating parameters on the reactor transient behavior during the accidents. The assumptions and initial conditions used for the RIA analysis in the Kori Unit 1 FSAR (Final Safety Analysis Report) are reexamined, and the corresponding analysis results are reassessed, based on the sensitivity analysis results, to be conservative and reliable.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.495-499
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• 1995

울진 1,2호기 RSTR 수행 시 울진 1,2호기 기존의 $\Delta$I Band를 해석하기 위해 RAOC방법을 적용 사고 해석을 수행하였다. 먼저 Xenon reconstruction model을 사용 축 방향 Xenon 분포를 생산한 다음, 정상 운전 상태와Condition ll상태에서 생산된 xenon 분포에 의한 축 방향 출력 분포를 사용 $F_{Q}$와 DNBR을 계산, Design Limit와 비교 만족하는 새로운 $\Delta$I band를 결정하였다. 새로운 band는 기존의 Design Limit의 변화를 주지 않으면서 울진 발전소 기존의 $\Delta$I band를 포함하면서 운전상의 유연성 창상을 기하게 되었다.$\Delta$I band를 포함하면서 운전상의 유연성 창상을 기하게 되었다.다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.543-548
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• 1996

이 연구의 목적은 고리 3/4호기 및 영광 1/2호기의 음의 중성자속 변화율에 의한 원자로 트립(NFRT) 설정치를 제거하여 불의의 제어봉 낙하 사고시에 원자로 트립을 방지하는 것이다. 현재의 인허가된 안전해석 방법론에 의하면 제어봉 낙하사고시에 NFRT에 의하여 원자로가 트립되고 결국 발전소의 이용율이 감소하게 되는데 본 연구에서 적용된 새로운 방법론으로는 이 NFRT 보호신호 없이도 제어봉 낙하사고시에 발전소의 안전성을 입증할 수 있다. 안전분석은 주기별로 다른 핵설계 자료 즉, 냉각재 온도상수, 전출력에서의 제어봉가 및 제어봉 삽입한계 등을 이용하여 수행되었다. 웨스팅하우스형 연료인 OFA 및 V5H에 대하여 고리 4호기 6주기외 3개 주기에 대하여 분석되었다. 분석된 주기들에 대해서 NFRT 신호 없이도 핵비등 이탈률(DNBR) 설계기준을 모두 만족하였다. 그러므로 고리 3/4호기 및 영광 1/2호기의 NFRT 신호는 제거할수 있고 이로써 제어봉 낙하사고시의 발전소 불시정지를 방지할수 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.562-567
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• 1995

The concept of VLDT (Variable Low DNBR Trip), a new CPC trip function, was proposed and applied to the events of increase in secondary heat removal, such as an excess feedwater event anti an IOSGADV (Inadvertent Opening S/G Atmospheric Dump Valve). Major assumption used in this study was no time delay to LOOP (Loss of Offsite Power) after turbine trip. In case of using this VLDT function, safety criterion of DNB would not be violated under the same condition as previous analysis without any change in thermal margin.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.550-555
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• 1995

국내 Westinghouse형 및 CE형 가압 경수로의 Non-LOCA 및 성능 분석을 수행할 수 있는 범용 전산 코드 TASS 1.0 코드를 한국원자력연구소에서 개발하였다. TASS 1.0의 노심 출력 계산은 Point Kinetics 모델과 1차원 확산 모델이 함께 내장되어 있어 축방향 출력 분포가 변하는 반응도 관련 사고 및 주증기관 파단 사고들에 대해서는 1차원 확산 모델을 사용하여 노심의 출력 계산이 가능하도록 개발되었다. 1차원 확산 모델의 적용 가능성 및 효과를 평가하기 위하여 Westinghouse형 발전소인 고리 3호기 7주기 및 CE형 발전소인 영광 3호기 1주기 전출력 제어봉 인출 사고에 대한 비교 분석 계산을 수행하였다. 비교 분석 계산 결과 1차원 확산 모델이 Point Kinetics 모델에 비해 DNBR 관점에서 보다 많은 운전 및 열적 여유도를 확보함이 판명되어 반응도 관련 사고 해석에서의 TASS 1.0 1차원 확산모델의 개선 효과를 입증하였다.

• 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).

• Journal of Institute of Control, Robotics and Systems
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• v.4 no.3
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• pp.400-405
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• 1998

본 논문에서는 한국형 원자력 발전소의 기준모델인 영광 3,4호기 운전원 훈련용 시뮬레이터의 모델링 절차와 ABB-CE 원전의 독특한 계통인 CPC/COLSS (Core protection Calculator/Core Operating Limit Supervisory System) 계통에 대한 모델링을 전개허고 있다. CPC/COLSS는 원자로를 포함하는 냉각재계통(NSSS)과 핵연료의 건전성을 보장하기위한 계통으로서 감시및 보호 과정에서의 계산을 디지털화시킴으로서 정확성과 함께 원자로의 안정성을 향상시킨 특색있는 계통이다. 따라서 영광 3,4호기 시뮬레이터에서는 CPC/COLSS 계통에 대한 정확한 모델링을 하여 시험을 통해 성능및 기능에 대한 검증을 마침으로서 CPC/COLSS 시뮬레이션 모델 개발이 성공적으로 되었고 영광 3,4호기 운전 특성에 맞는 시뮬레이터를 개발하였다.