• Nuclear Engineering and Technology
• /
• 제55권8호
• /
• pp.3088-3101
• /
• 2023

A pressurized vessel-pipe-safety valve (PVPSV) combination is a commonly used configuration in nuclear power plants, and a good numerical model is essential for the system design, sizing and performance optimization. However, owing to the large-scale and cross-scale features, it is still a challenge to build a system level numerical model with both high accuracy and efficiency. To overcome this, a novel system level modeling method which can synthesize the advantages of various models is proposed in this paper. For system modeling, the analytical approach, the method of characteristics (MOC) and the surrogate model approach are respectively adopted to predict the dynamics of the pressure vessel, the connecting pipe and the safety valve, and different models are connected through data interfaces. With this system model, dynamic simulations were carried out and both the stable and the unstable system responses were obtained. For the model verification purpose, the simulation results were compared with those obtained from experiments and full CFD simulations. A good agreement and a better efficiency were obtained, verifying the ability of the model and the feasibility of the modeling method proposed in this paper.

• 한국태양에너지학회:학술대회논문집
• /
• 한국태양에너지학회 2009년도 추계학술발표대회 논문집
• /
• pp.409-414
• /
• 2009

Since cryogens are stored at very low temperatures, the cryogenic storage systems are quite sensitive to heat leaks. Even though the vessel operated under sealed condition with vacuum insulation and reflective coatings are used, the heat leakage into the vessel is still unavoidable. Therefore, this paper concerns with numerical study of self-pressurization used to analysis the optimum design with the variation volume fraction, effect of heat flux and storage pressure of liquid nitrogen. The result shows that as the volume fraction increases, the pressure rise reduces and the relatively at atmosphere pressure is better than the higher one. In addition, higher heat flux leads the pressure rise increases faster than low one. The additional of heat pipe system to reduce the pressure rise rate also has been done. By this comparison, the optimum design for storage umbilical cord blood can be selected.

• Nuclear Engineering and Technology
• /
• 제27권5호
• /
• pp.710-720
• /
• 1995

원자로에서 펌프에 의해 야기되는 맥동 압력은 원자로 내부 구조물에 진동과 손상을 줄 수 있기 때문에 관심이 증가되고 있다. 본 연구에서는 냉각관과 환형관(원자로 압력 용기와 노심 보호 지지대 사이)으로 구성된 기하 형태에서 펌프에 의해 야기되는 맥동 압력을 해석할 수 있는 수력학적 모델을 개발하였다. 수학적 지배 방정식은 압축성, 비점성 유체에 대해 선형화된 Navier-Stokes 방정식이다. 냉각관과 환형관을 따로 분리하여 해석하고 두영역의 커플링 영향을 고려하였다. 또한 본 기하 형태에서 펌프맥동 압력에 영향을 미치는 주요 기하 인자에 대한 평가를 수행하였다. 본 해석 결과와 실험차를 비교하여 만족할 만한 결과를 얻었다.

• 한국압력기기공학회 논문집
• /
• 제12권1호
• /
• pp.101-106
• /
• 2016

NSSS (Nuclear Steam Supply System) and BOP (Balance of Plant) design works for PGSFR (Prototype Gen-IV Sodium Fast Reactor) have been conducted in Korea. NSSS major components, e.g. reactor vessel, steam generator and secondary sodium main pipes, are designed according to the rule of ASME boiler and pressure vessel code division 5, in which DBA (Design by Analysis) methods are used in the stress assessments. However, there is little discussions about detail rules for BOP piping design. In this paper, the detail methodologies of BOP piping stress assessment are discussed including safety systems and non-safety system pipings. It is confirmed that KEPIC MGE(ASME B31.1) and ASME BPV code division 5 HCB-3600 can be used in stress assessments of non-safety pipes and class B pipes, respectively. However, class A pipe design according to ASME BPV code division 5 HBB-3200 has many difficulties applying to PGSFR BOP design. Finally, future development plan for class A pipe stress assessment method is proposed in this paper.

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

• 한국전산구조공학회:학술대회논문집
• /
• 한국전산구조공학회 1996년도 봄 학술발표회 논문집
• /
• pp.27-39
• /
• 1996

배관을 구속시키기 위한 원통형 배관 지지대 (Trunnion Pipe Support)가 부착된 배관의 응력해석 을 위하여 유한요소해석을 사용하였다. 해석결과로 부터 얻어진 응력은 두께에 대한 평균 및 선 형 응력으로 분류 되었으며, 분류된 응력값은 압력에 의한 일차응력계수(B$_1$)와 이차응력계수(C$_1$), 모멘트에 의한 일차응력계수(B$_2$)와 이차응력계수(C$_2$)를 추정하기 위하여 ASME Code에 정의된 것과 일치하게 해석되었다. 무차원의 함수로써 응력계수에 대한 경험식을 개발하기 위하여 여러 모델의 해석을 수행하였다.

• International Journal of Korean Welding Society
• /
• 제1권1호
• /
• pp.1-6
• /
• 2001

This paper describes several advances in sensor and process control techniques for applications in Submerged Arc Welding (SAW), which combine to give a fully automatic system capable of controlling and adapting the overall welding process. This technology has been applied in longitudinal and spiral pipe mills and in pressure vessel production.

• Journal of Advanced Marine Engineering and Technology
• /
• 제38권10호
• /
• pp.1244-1250
• /
• 2014

The temperature of the main engine cabin of commercial vessel is very high. The material SS-316L undergoes creep damage at temperatures exceeding $450^{\circ}C$. It is essential to maintain the highly stressed engine cabin below the creep regime. Hence, seawater is employed in this kind of maritime vehicles as cooling liquid. It obtains the thermal energy at the cooling pipe line after passing through main engine cooling system. To harness the energy in the seawater, a turbine can be installed to absorb the energy in the seawater before being released into the sea. In this study, a cooling pipe line is selected to apply the tubular type hydro turbine for transferring the energy. Numerical analysis for investigating the performance and the internal flow characteristics of the tubular turbine is conducted. The results show that the maximum efficiency of 85.8% is achieved although the efficiency drops rapidly at partial flow rate condition. The efficiency descends slowly at the condition of excess flow rate. There is a relatively wide operating range of flow rate of this turbine to keep high efficiency at the excess flow rate condition. For the internal flow of the turbine, there is uniform streamline on the suction and pressure sides of the blade at the design point. However, the secondary flow appears at the suction and pressure sidesat the excess flow rate.In addition, it appears only at pressure side at the partial flow rate condition.

• 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.

• 한국정밀공학회:학술대회논문집
• /
• 한국정밀공학회 2000년도 추계학술대회 논문집
• /
• pp.547-550
• /
• 2000

The steel pipe for fluid catalytic cracking(FCC) unit. petroleum refinery, is lined with refractory to protect the system from high-temperature of the internal flow. The property of the refractory has an effect upon the stress analysis of FCC unit. Because 1-D pipe element or 3-D shell element are usually used in commercial codes of stress analysis to evaluate the structural soundness, the equivalent elastic modulus considering steel and refractory should be applied. In the research, the theoretical method to obtain the value of the equivalent property is introduced and then the stress analysis is carried out with the part of FCC unit.