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Performance Analysis of a Once-Through Steam Generator for Constant Thermal Power Operation
Hun Sik Han(한훈식),Juhyeon Yoon(윤주현),Young In Kim(김영인) 대한기계학회 2017 대한기계학회 춘추학술대회 Vol.2017 No.11
A numerical study is made of the operation strategy of a once-through steam generator considering degradation in the thermal-hydraulic performance. The performance of the steam generator is gradually degraded as it experiences fouling, clogging, and plugging during its lifetime. To maintain the thermal power of the steam generator, therefore, its operating condition should be properly changed according to the degradation rate. Emphasis is given to the operation strategy of the steam generator to transport a constant thermal power regardless of the degradation rate. Comprehensive numerical solutions to the governing equations are acquired. It is found that a constant thermal power can be maintained by properly adjusting the secondary coolant outlet pressure with a variation of the superheat degree and secondary coolant pressure drop. A constant thermal power operation curve is obtained, and the thermal-hydraulic performance of the steam generator is analyzed according to the degradation rate in terms of the degree of superheat, secondary coolant pressure drop, temperature distribution, and quality distribution in the tube.
일체형원자로 증기발생기 카세트 하단에 설치된 오리피스의 최적설계 연구
강형석(Hyung Seok Kang),윤주현(Juhyeon Yoon),김환열(Hwan Yeol Kim),조봉현(Bong Hyun Cho),이두정(Doo Jeong Lee) 한국전산유체공학회 1998 한국전산유체공학회 학술대회논문집 Vol.1998 No.-
A new advanced integral reactor of 330 MWt capacity named SMART(System-integrated Modular Advanced Reactor) is currently under development at KAERl(Korea Atomic Energy Research Institute). One of the major design features of the integral reactor is locating the steam generators(SG) inside reactor vessel and eliminating the possibility of LB LOCA(large Break Loss of Coolant Accident). Orifices are fitted at the low part of steam generator cassette to stabilize and balance coolant flow distribution in the MCP (Main Circulation Pump) pressure header. A sensitivity analysis is conducted to determine the optimum orifice size using computer code "CFX".
Minkyu Lee(이민규),Seungyeob Ryu(유승엽),Juhyeon Yoon(윤주현),Young In Kim(김영인) 대한기계학회 2017 대한기계학회 춘추학술대회 Vol.2017 No.11
A series of transient analyses related to overpressure protection of an integral reactor is conducted to investigate the steam pressurizer (PZR) characteristic. The transient is considered the loss of load accident with delay reactor trip, and it yields thermally induced surge flow into PZR. The in-surged flow leads the excessive pressurization, and the system shall be protected from the overpressure by the Pressurizer Safety Valve (PSV). Mixing model and two phase depressurization model are considered to evaluate the PZR pressure behavior during the transient. Critical flow model is considered to be developed at throat of PSV when the saturated steam is discharged through the PSVs. Numerical analysis results show that the reactor can be provided overpressure protection with the PSV opening. Additionally, the flashing phenomenon is found during the depressurization, and the depressurization rate is reduced by the phase change effect after the onset of flashing.
연구용 원자로의 다양한 운전범위에 따른 일차계통펌프 설계
서경우(Kyoungwoo Seo),윤현기(Hyungi Yoon),지대영(Dae-young Chi),윤주현(Juhyeon Yoon) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.10
A design of a primary cooling system, which removes the heat generated from the core, is dependent on a purpose of a research reactor. A reactor can require variable operating ranges for the primary cooling system. To operate the variable operating range, the primary cooling pump is designed with an inverter or a control valve in the system. The system pressure losses are calculated by the analytic and the numerical methods for the required flow rates estimated from the required core powers and temperature differences. It can be designed that the pump curve is moved by the rpm control with an inverter for the variable pump operating range. If a control valve is installed in the system, a pump curve can be used for the variable range.
Hun Sik Han(한훈식),Han-Ok Kang(강한옥),Juhyeon Yoon(윤주현),Young In Kim(김영인),Youngmin Bae(배영민),Sang Ji Kim(김상지) 대한기계학회 2019 대한기계학회 춘추학술대회 Vol.2019 No.11
A numerical study is conducted for the secondary side screw-type tube inlet orifice design of a once-through steam generator. Various tube plugging conditions and power levels are considered, and the secondary coolant inlet temperature is adjusted to maintain a constant thermal power. Comprehensive numerical solutions are acquired to evaluate the minimum orifice length under various operating conditions, and the required minimum orifice length to suppress the flow oscillation below the allowable level is evaluated. The results obtained show that the lowest power level results in the highest minimum orifice length and non-plugging condition provides a limiting case for the orifice length criterion.
Effect of DVI Nozzle Location on the Thermal Mixing in the RVDC
강형석(Hyung Seok Kang),조봉현(Bong Hyun Cho),김환열(Hwan Yeol Kim),윤주현(Juhyeon Yoon),배윤영(Yoon Yeong Bae) 한국전산유체공학회 1998 한국전산유체공학회지 Vol.3 No.1
한국형 차세대원자로에서는 비상노심 안전주입수가 저온관을 통하지 않고 원자 로용기에 직접 주입된다. 원자로용기의 가압열충격과 열수력적 관점에서 최적의 노즐위치를 결정하기 위해서 전산유체역학을 활용하였다. 상용 전산유체코드인 CFX를 이용하여 원자로 하향유로를 모사하는 해석대상 격자를 다중블록으로 형성한 다음 유동장을 비압축성 Navier-Stokes 운동량 방정식, 에너지 방정식과 표준 k-ε 난류모형 등으로 모형화하여 3차원 비정상상태 계산을 수행하였다. CFX에서는 경계 밀착좌표계 - 비엿물림격자와 SIMPLE 알고리즘을 사용한다. 본 연구결과 원자로용기의 가압열충격 관점에서 가장 보수적인 사고인 증기관 과단사고시에도 열적혼합이 잘 일어나 가압열충격이 발생할 가능성이 없는 것으로 판단되며 안전주입수 노즐이 저온관 바로 위에 위치할 때 원자로 하향유로 내의 온도 분포가 가장 균일하여 열적 혼합 관점에서는 최적의 위치로 판단된다.
Effect of DVI Nozzle Location on the Thermal Mixing in the RVDC
강형석(Hyung Seok Kang),조봉현(Bong Hyun Cho),김환열(Hwan Yeol Kim),윤주현(Juhyeon Yoon),배윤영(Yoon Yeong Bae) 한국전산유체공학회 1998 한국전산유체공학회지 Vol.3 No.1
한국형 차세대원자로에서는 비상노심 안전주입수가 저온관을 통하지 않고 원자 로용기에 직접 주입된다. 원자로용기의 가압열충격과 열수력적 관점에서 최적의 노즐위치를 결정하기 위해서 전산유체역학을 활용하였다. 상용 전산유체코드인 CFX를 이용하여 원자로 하향유로를 모사하는 해석대상 격자를 다중블록으로 형성한 다음 유동장을 비압축성 Navier-Stokes 운동량 방정식, 에너지 방정식과 표준 k-ε 난류모형 등으로 모형화하여 3차원 비정상상태 계산을 수행하였다. CFX에서는 경계 밀착좌표계 - 비엿물림격자와 SIMPLE 알고리즘을 사용한다. 본 연구결과 원자로용기의 가압열충격 관점에서 가장 보수적인 사고인 증기관 과단사고시에도 열적혼합이 잘 일어나 가압열충격이 발생할 가능성이 없는 것으로 판단되며 안전주입수 노즐이 저온관 바로 위에 위치할 때 원자로 하향유로 내의 온도 분포가 가장 균일하여 열적 혼합 관점에서는 최적의 위치로 판단된다.
Static Instability of a Once-Through Steam Generator with a Modular Feedwater Line
Jae-Kwang Seo(서재광),Han-Ok Kang(강한옥),Juhyeon Yoon(윤주현),Keung-Koo Kim(김긍구) 대한기계학회 2006 대한기계학회 춘추학술대회 Vol.2006 No.11
Static flow instability of a once-through steam generator (OTSG) with a modular feedwater line is an instability related to the change of a flow direction in individual steam generating U-shaped channels operating at a given pressure difference. The nature of the static instability is close to a Ledinegg instability [1] related to the presence of multiple flows at the full hydraulic curve of a U-shaped channel. In this paper, the conditions for a reverse flow for the OTSG with a U-shaped modular feedwater line (MFL) are studied. From the results of the studies, it is revealed that the change of a flow direction in the MFL is due to a boiling of the feedwater in the downcomer branch of the U-shaped MFL and that multiple flows start in an area of the extremes corresponding to the minimum pressure difference of the hydraulic curves. Calculation models for predicting a threshold of a static instability for the OTSG of interest is proposed and the analysis results are compared with the experimental data.