• 한국유체기계학회 논문집
• /
• 제17권6호
• /
• pp.109-114
• /
• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• Nuclear Engineering and Technology
• /
• 제38권2호
• /
• pp.185-194
• /
• 2006

Transient operations for an integral type reactor, SMART-P, have been experimentally investigated using a thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), in order to verify the system design and performance of the SMART-P, a pilot plant of SMART. The VISTA facility was subjected to various accident conditions such as feedwater increase and decrease, loss of coolant flow, and control rod withdrawal accidents in order to elucidate the thermal-hydraulic responses following such accidents and finally to verify the system design of the SMARTP. Full functional control logics have been implemented in the VISTA facility in order to control the required control action for an accident simulation. As one of the sensitivity tests to verify the PRHRS performance, the effects of the initial water level in the compensation tank are experimentally investigated. When the initial water level is 16%, the water is quickly drained and nitrogen gas is then introduced into the PRHR system, resulting in deterioration of the PRHRS performance. It is thus found that nitrogen ingression should be prevented to ensure stable PRHRS operation.

• Nuclear Engineering and Technology
• /
• 제19권1호
• /
• pp.10-21
• /
• 1987

극한적인 열제거 기능 상실사고인 급수관 파열사고에 대한 불확실성 해석을 반응표면방법과 Monte Carlo모사를 이용해서 원자력 1호기에 대하여 수행하였다. 여러번의 RELAP4/MOD6를 이용한 급수관 파열사고 해석을 통해 불확실성 해석의 Data Base를 마련하였으며, 비교 목적으로 평가모형 계산도 수행하였다. 급수관 파열사고 이후의 원자로 냉각재계통 최대 압력에 미치는 영향을 조사비교하기 위해 2증류의 입력 Set에 대한 반응표면방법이 활용되었다. 첫 Set는 6개의 주요 발전소 운전변수로 구성되며, 둘째 Set는 5개 주요 모형변수로 구성된다 결과의 비교 분석을 통해 모형변수의 불확실성 이 최대 압력에 미치는 영 향이 운전변수 불확실성의 영향보다 매우 큰 것이 밝혀졌고, 최대 압력 증가의 약 9%에 해 당되는 여유도 개선도 확인되었다. 또한, 평가모델에서 인정되고 있는 초기 냉각재 노심입구 온도에 대한 가정은 잘못된 것으로 밝혀졌다.

• 멤브레인
• /
• 제29권6호
• /
• pp.314-322
• /
• 2019

기후변화로 인해 가속화되고 있는 충남 지역 가뭄현상에 대응하고 대산 임해산업단지의 증가하는 용수 수요를 만족하기 위해서, K-water에서는 100,000 ㎥/일 규모 역삼투 해수담수플랜트 건설 사업을 추진하고 있다. 이에 본 연구에서는 해수담수플랜트 운영비용의 70% 이상을 담당하는 역삼투 공정 성능에 영향을 미치는 주요 인자에 대한 성능 분석을 수행하였다. 대산 지역 해수 염분농도 및 수온 변화 조건에서 RO 공정의 전력소모는 2.39 ± 0.13 kWh/㎥로 나타났으며, 막여과유속과 회수율이 낮을수록 전력소모가 절감되어 연간 운영비용이 감소하였다. 주요 막 제조사별 고유량 막의 성능 비교 결과, 전량 2단 여과공정(full two pass) 기준 생산수 TDS는 평균 3.84 mg/L로 양호하였고, 전력소모는 2.22 ± 0.13 kWh/㎥ 수준으로 확인되었다. 역삼투 공정 구성을 전량 2단 여과방식에서 partial 또는 split partial 방식으로 변경함에 따라 전력소모는 최대 0.29 kWh/㎥, 막모듈 설치비용은 최대 15.6% 절감 가능할 것으로 기대된다.

• Corrosion Science and Technology
• /
• 제12권3호
• /
• pp.125-131
• /
• 2013

The most common pipe wall thinning degradation mechanisms that can occur in the steam and feedwater systems are FAC (Flow Acceleration Corrosion), cavitation, flashing, and LDIE (Liquid Droplet Impingement Erosion). Among those degradation mechanisms, FAC has been investigated by many laboratories and industries. Cavitation and flashing are also protected on the piping design phase. LDIE has mainly investigated in aviation industry and turbine blade manufactures. On the other hand, LDIE has been little studied in NPP (Nuclear Power Plant) industry. This paper presents the development of prediction system for pipe wall thinning caused by LDIE in terms of erosion rate based on air-water ratio and material. Experiment is conducted in 3 cases of air-water ratio 0.79, 1.00, and 1.72 using the three types of the materials of A106B, SS400, and A6061. The main control parameter is the air-water ratio which is defined as the volumetric ratio of water to air (0.79, 1.00, 1.72). The experiments were performed for 15 days, and the surface morphology and hardness of the materials were examined for every 5 days. Since the spraying velocity (v) of liquid droplets and their contact area ($A_c$) on specimens are changed according to the air-water ratio, we analyzed the behavior of LDIE for the materials. Finally, the prediction equations(i.e. erosion rate) for LDIE of the materials were determined in the range of the air-water ratio from 0 to 2%.

• 대한환경공학회지
• /
• 제35권3호
• /
• pp.200-205
• /
• 2013

전처리 공정은 막의 오염 정도를 제어할 수 있는 유일한 방법이다. 막의 오염 현상은 피할 수 없는 중요한 문제이며 RO 공급수로 적합한 전처리 공정의 선택이 중요하다. 본 연구는 pH, 해수의 탁도, 수온, 응집제 주입량 그리고 SDI 측정 막재질과 같은 SDI 값에 영향을 미치는 인자들을 평가 하기위해 수행되었다. 그 결과 해수의 탁도는 여과수의 SDI 값에 약간의 영향을 미친것으로 조사되었다. 0.45 um 공극 크기를 가지는 SDI 측정 막은 소수성 막 보다 동일한 재질의 친수성 막을 이용하는 것이 분석 결과의 신뢰성과 재현성을 확보할 수 있었다. pH 7.0 이하의 조건에서 pH가 감소할수록 SDI 값은 증가한 것으로 나타났다. 그리고 수온은 SDI 값은 큰 영향을 미친것으로 조사되었다.

• 멤브레인
• /
• 제4권4호
• /
• pp.213-220
• /
• 1994

정유산업 방출폐수를 재처리하여 양질의 공업용수로 활용하고자 역삼투막 및 공정으로 구성된 재활용 system을 고안하여 현장 pilot 실험을 수행하였다. 생산공정 운전상황에 따라 불규칙적으로 변하는 폐수 특성에도 불구하고 역삼투막의 경우 10~15$kg/cm^2$의 운전압력 범위내에서 96~99%정도의 비교적 높은 염배제율을 보였으며 본 실험을 통하여 얻은 재생수의 경우 냉각탑 공급용수로써의 충분한 가능성을 보여 주었다. 그러나 본 연구에서 제안 사용된 여과형 전처리 공정만으로는 충분한 처리효율을 기대하기 어려웠으며, 이는 각 여과공정법 처리수의 수질분석 결과 및 NaOH를 이용한 역삼투막 세정 결과로부터 쉽게 확인되었다.

• Nuclear Engineering and Technology
• /
• 제52권12호
• /
• pp.2699-2708
• /
• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• Nuclear Engineering and Technology
• /
• 제53권5호
• /
• pp.1436-1445
• /
• 2021

To obtain the dynamic characteristics of reactor secondary circuit under transient conditions, the system analysis program was developed in this study, where dynamic models of secondary circuit were established. The heat transfer process and the mechanical energy transfer process are modularized. Models of main equipment were built, including main turbine, condenser, steam pipe and feedwater system. The established models were verified by design value. The simulation of the secondary circuit system was conducted based on the verified models. The system response and characteristics were investigated based on the parameter transients under emergency shutdown and overload. Various operating conditions like turbine emergency shutdown and overspeed, condenser high water level, ejector failures were studied. The secondary circuit system ensures sufficient design margin to withstand the pressure and flow fluctuations. The adjustment of exhaust valve group could maintain the system pressure within a safe range, at the expense of steam quality. The condenser could rapidly take out most heat to avoid overpressure.

• 한국안전학회지
• /
• 제36권3호
• /
• pp.74-80
• /
• 2021

Korea Atomic Energy Research Institute (KAERI) has been operating an integral effects test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), according to APR1400 for transient experimental and design basis accident simulation. Moreover, based on the experimental data, the domestic standard problem (DSP) program has been conducted in Korea to validate system codes. Recently, through DSP-05, the performance of the passive auxiliary feedwater system (PAFS) in the event of multiple steam generator tube rupture (MSGTR) has been analyzed. However, some errors exist in the reference input model distributed for DSP-05. Furthermore, the calculation results of the heat loss correlation for the secondary system presented in the technical report of the reference indicate that a large difference is present in heat loss from the target value. Thus, in this study, the reference model is corrected using the geometric information from the design report and drawings of ATLAS. Additionally, a new heat loss correlation is suggested by fitting the results of the heat loss tests. Herein, MSGTR-PAFS accident analysis is performed using MARS-KS 1.5 with the improved model. The steady-state calculation results do not significantly differ from the experimental values, and the overall physical behavior of the transient state is properly predicted. Particularly, the predicted operating time of PAFS is similar to the experimental results obtained by the modified model. Furthermore, the operating time of PAFS varies according to the heat loss of the secondary system, and the sensitivity analysis results for the heat loss of the secondary system are presented.