• Nuclear Engineering and Technology
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• 제33권4호
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• pp.441-456
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• 2001

ABB-CE＇s regionalization methodology for the core inlet flow distribution map is reviewed. This methodology merges the test data of fuel assembly locations which are either in symmetry or strongly correlated with others. It increases the number of available test data for each regional flow factor It makes up effectively for the deficiency due to limited number of test data. It also contributes to making the core inlet flow distribution smoother not only locally but also over the entire core, and to relieving the impacts of test errors that may happen due to some do- calibrated local pressure measurement taps. As a result, the core inlet How distribution data becomes more statistically useful and thus the conservatism involved in handling the core inlet flow factors for the thermal margin analysis is expected to be reduced. Meanwhile, the regionalized map may lose the unique local characteristics in core inlet flow distribution too much. By an alternative approach introduced in the present work, it is shown that such a disadvantage can be mitigated somewhat if the engineering judgement is made more

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Journal of Advanced Marine Engineering and Technology
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• 제39권6호
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• pp.604-608
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• 2015

본 연구에서는 다심관(Multi Core Tube)에 단열재로 글라스울(Glass wool)을 사용한 단열 다심관(Insulated Multi Core Tube)의 열전달 특성에 관하여 검토하였다. 제작된 단열 다심관에 대하여 외기온도, 유압 오일 공급온도, 유압 오일 공급유량에 따른 단열 다심관 내부의 유압 오일의 온도특성에 관하여 실험 및 모델링을 통하여 검토하였다. 그 결과 본 연구의 범위 내에서 최소 유압 오일 공급유량인 0.29(l/min)인 경우 실험결과와 수치해석 결과의 온도차이가 최대 약 $3^{\circ}C$정도 발생하였다. 외기온도가 일정한 경우 유압 오일 공급온도가 높을수록 유압 오일의 공급유량에 관계없이 유압 오일 출구온도가 높아지고, 유압 오일 공급유량이 1.01(l/min)이상일 경우에는 유압 오일의 온도강하에 외기온도의 영향이 적음을 알 수 있었다.

• 한국수자원학회논문집
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• 제52권4호
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• pp.235-243
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• 2019

급격한 기상변화로 인한 극한 강우와 집중폭우의 발생빈도 증가로 기존 수방시설의 한계 용량을 초과해 도심지 침수피해가 빈번하게 발생하고 있다. 최근 도시화 추세가 급격하게 빨라지면서 수방시설 등 사회기반시설에 대한 지하공간 개발의 필요성이 증가하고 있으며, 지하공간을 활용한 지하방수로와 지하저류지 기술이 급부상하고 있다. 본 연구에서는 지하유입시설의 대표적 형상인 접선식 유입구와 나선식 유입구에 대한 공기 배출효과를 분석하기 위해 유입유량 변화에 따른 수직갱 내부 공기공동(air-core)의 형상 크기를 계측했다. 나선식 유입구의 경우, 저유량 유입조건에서 와류 유도 효과를 개선하기 위해 유입부 바닥면에 계단형 다단식 구조를 도입했다. 수직갱 내부 공기공동의 전체적인 평균 단면적의 경우, 다단식 나선 유입구가 접선식 유입구보다 10% 정도 크게 나타나 고유량 유입조건에서 높은 공기 배출 효과와 유입효율을 나타냈다. 접선식 유입구의 경우, 유입구가 가지는 고유 성능을 유지할 수 있는 최대 유량 조건을 초과하면서 공기 배출 효과가 감소하기 시작했다. 또한, 실험에서 사용된 접선식 유입구와 다단식 나선 유입구 모형에 활용 가능한 기초자료를 제공하기 위해 수직갱 내부 위치에 따른 공기공동 단면적에 대한 실험식(empirical formula)을 제시했다.

• 한국전산유체공학회지
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• 제20권1호
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• pp.84-91
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• 2015

SMART adopts, very unique facility, an FMHA to enhance the thermal and flow mixing capability in abnormal conditions of some steam generators or reactor coolant pumps. The FMHA is important for enhancing thermal mixing of the core inlet flow during a transient and even during accidents, and thus it is essential that the thermal mixing characteristics of flow of the FMHA be understood. Investigations for the mixing characteristics of the FMHA had been performed by using experimental and CFD methods in KAERI. In this study, the temperature distribution at the core inlet region is investigated for several abnormal conditions of steam generators using the commercial code, FLUENT 12. Simulations are carried out with two kinds of FMHA shapes, different mesh resolutions, turbulence models, and steam generator conditions. The CFD results show that the temperature deviation at the core inlet reduces greatly for all turbulence models and steam generator conditions tested here, and the effect of mesh refinement on the temperature distribution at the core inlet is negligible. Even though the uniformity of FMHA outlet hole flow increases the thermal mixing, the temperature deviation at the core inlet is within an acceptable range. We numerically confirmed that the FMHA applied in SMART has an excellent mixing capability and all simulation cases tested here satisfies the design requirement for FMHA thermal mixing capability.

• Nuclear Engineering and Technology
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• 제29권3호
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• pp.186-195
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• 1997

When asymmetric thermal-hydraulic conditions occur between cold legs, the core inlet temperature will be nonuniform if the coolant is not mixed perfectly in the lower plenum. These uneven core inlet conditions may induce the change in core power distribution. Thus realistic prediction of thermal mixing is important in such abnormal conditions. In this study, reactor internals, which are scaled down as to conserve the flow area ratio, are set up in the model of KORI Unit 1 with the scaling factor of 1/710 by volume and coolant temperatures are measured beneath the lower core plate. Based on experimental results, the ability of COMMIX-1B code to simulate the coolant mixing phenomena in the lower plenum is estimated. The results show that complete mixing never occurs in any conditions and the mixing pattern is characterized according to the plant type.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.435-440
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• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.71-81
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• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1995년도 춘계학술발표회 초록집
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• pp.113-116
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• 1995

Simplified approach has been adopted for the prediction of the thermal behavior of CANDU reactor core during power transients. Based on the assumption that the ratio of mass flow rate for each core channel does not vary during the transient, quasy-steady state analysis technique is applied with predicted core inlet boundary conditions(total mass flow rate and specific enthalpy). For restricted transient case, the presented method shows functionally reasonable estimation of core thermal behavior which could be implemented in the fast running reactor simulation program.

• Nuclear Engineering and Technology
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• 제15권4호
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• pp.236-247
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• 1983

한국원자력 1호기(KNU-1)의 설계 및 운전자료를 이용하여 가압경수로 운전변수들의 변화에 대한 DNBR의 민감도를 분석하였다. 본 민감도 분석에는 원자로 출력, 압력, 냉각수 주입유량, 냉각수 주입온도, 반경방향 및 축방향 출력분포 그리고 축방향 출력편차 등의 운전변수가 고려되었다. 민감도 분석을 위하여는 노심의 열수력 해석용 전산코드인 COBRA-IV-K를 사용하였는데 본 코드는 COBRA-IV-i의 수정판으로써 한국에너지연구소에서 일부 프로그램을 수정하였고 또한 신뢰도도 확인하였다. 민감도 분석을 수행하기 전에 KNU-1 원자로심의 설계 및 운전조건을 근거로 하여 기초 계산을 수행하고 이 결과를 본 민감도 분석의 기본자료로 삼았다. 민감도 분석결과 원자로의 DNBR 열설계에 있어서 가장 민감한 운전변수는 냉각수 주입온도이고 가장 둔감한 변수는 축방향 출력분포라는 것이 밝혀졌다.