• Title/Summary/Keyword: 정지봉 장치

Search Result 9, Processing Time 0.029 seconds

Analysis of the Momentary Interruption Impact on the HANARO Operation (순간정전이 하나로 운전에 미치는 영향 분석)

  • Kim, Hyeong-Gyu;Jeong, Hwan-Seong;Choe, Yeong-San;U, Jong-Seop
    • Proceedings of the Korean Nuclear Society Conference
    • /
    • 2004.10a
    • /
    • pp.655-656
    • /
    • 2004
  • 1) 제어봉의 전자 클러치는 직류전원 공급 장치에 의해 12V의 직류 전력을 공급받으며 전압 강하에 대한 내성이 좋다. 제어봉은 10V의 전압이 525msec 이상 지속될 때 전자력 상실로 낙하한다. 완전정전(0V)이 발생하여도 직류전원 공급 장치는 500msec 동안 전자클러치에 직류전력을 공급하여 제어봉의 연결 상태를 유지하도록 한다. 2) 정지봉 계통에 대한 전압강하의 영향은 제어회로를 구성하는 전자접촉기의 개방에 의하여 펌프의 전원공급이 차단되고, 그 결과 정지봉이 낙하한다. 정지봉은 펌프의 전원이 상실되면 수압 실린더의 압력 상실로 약 1000msec 후에 낙하한기 시작한다. 그림 2는 제어봉 및 정지봉에 대한 정전 영향을 시간에 따라 표시한 것이다. 3) 1차 및 2차 냉각계통의 부족전압 계전기에 의해 펌프가 정지할 때까지 저유량 신호 및 N/T mismatch 신호에 의한 원자로 정지신호는 발생되지 않는다. 따라서 정지봉 및 제어봉 계통에 적용하고자 하는 순간정전 보상장치는 부족전압 계전기 동작시간 이내의 보상시간에서만 가능할 것이다.

  • PDF

CRDM내 이동 권선 신호를 이용한 고리2호기 제어봉 낙하 시간 측정 시험

  • 윤명현;김기훈;신창훈
    • Proceedings of the Korean Nuclear Society Conference
    • /
    • 1998.05a
    • /
    • pp.258-263
    • /
    • 1998
  • 원자력 발전소의 제어봉 낙하 시간 측정 시험은 원자로 비상 정지 명령에 따른 제어봉의 낙하 속응성이 지침서의 제한치 이내에 있는지를 주기적으로 확인함으로써 제어봉의 원자로 안전 운전 및 정지 기능을 확인하는 중요한 시험이다. 현재 고리2호기를 비롯한 WH형 발전소들은 제어봉이 디지틀 제어봉 위치 지시(DRPI) 코일속을 낙하할 때 유기되는 전압을 검출하여 제어봉 낙하 시간을 측정하는 방법을 사용하고 있다. 이 방법은 제어봉을 뱅크별로 인출한 후 하나씩 개별 낙하시키기 때문에, 제어봉 낙하 시간 측정 시험에 많은 시간과 다수의 인력이 소요되며, 측정시 격납용기내에 작업자가 장시간 체류해야 하는 단점이 있다. 이러한 DRPI 신호 이용 제어봉 낙하 시간 측정 방법의 단점을 보완하여 제어봉 구동 장치(CRDM)내 이동권선 신호를 이용하는 다중 낙하 방식의 새로운 측정 방법을 제안하고, 측정 장치를 개발하였다. 개발된 제어봉 낙하 시간 측정 장치를 이용하여, 고리2호기 제어봉 낙하 시간 측정 시험시 기존의 방법과 병행 측정 시험을 수행하고 그 결과를 비교 검토하여 충분히 정확하게 제어봉 낙하 시간을 측정할 수 있음을 보였다.

  • PDF

Performance test and factor analysis on the performance of shutoff units with the research reactor (연구용 원자로의 정지봉 장치 성능에 미치는 인자 분석과 성능 시험)

  • Kim, Kyoung-Rean;Kim, Seoug-Beom;Ko, Jae-Myoung;Moon, Gyoon-Young;Park, Jong-Ho
    • The KSFM Journal of Fluid Machinery
    • /
    • v.10 no.2 s.41
    • /
    • pp.41-45
    • /
    • 2007
  • The shutoff unit was designed to provide rapid insertion of neutron absorbing material into the reactor core to shutdown the reactor quickly and also to withdraw the absorber slowly to avoid a log-rate trip. Four shutoff units were installed on the HANARO reactor but the half-core test facility was equipped with one shutoff unit. The reactor trip or shutdown is accomplished by four shutoff units by insertion of the shutoff rods. The shutoff rod(SOR) is actuated by a directly linked hydraulic cylinder on the reactor chimney, which is pressurized by a hydraulic pump. The rod is released to drop by gravity, when triplicate solenoid valves are de-energized to vent the cylinder. The hydraulic pump, pipe and air supply system are provided to be similar with the HANARO reactor. The shutoff rod drops for 647mm stroke within 1.13 seconds to shut down the reactor and it is slowly inserted to the full down position, 700mm, with a damping. We have conducted the drop test of the shutoff rod in order to show the performance and the structural integrity of operating system of the shutoff unit. The present paper deals with the 647mm drop time and the withdrawal time according to variation of the pool water temperature, the water level and the core flow.

Seismic Test of the Control Rod Drive Mechanism for JRTR (JRTR 제어봉구동장치의 내진시험)

  • Choi, Myoung-Hwan;Kim, Gyeong-Ho;Sun, Jong-Oh;Cho, Yeong-Garp
    • Transactions of the Korean Society for Noise and Vibration Engineering
    • /
    • v.26 no.5
    • /
    • pp.552-558
    • /
    • 2016
  • A control rod drive mechanism(CRDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod within a reactor core to control the reactivity of the core. The CRDM for Jordan Research and Training Reactor with 5MW power has been designed and fabricated based on the HANARO’s experience through KAERI and DAEWOO consortium. This paper describes the seismic test results to demonstrate the operability, the drop performance and the structural integrity of CRDM during or after seismic excitations. The seismic tests are carried out under 5 OBE and 1 SSE loads at three Test Rigs simulating the reactor structure and the pool top. From the tests, the CRDM is smoothly driven without a malfunction of stepping motor under OBE load. The pure drop time under OBE and SSE loads is measured as 1.169s and 1.855s to meet the design requirement. Also, it is found that the CRDM maintains the structural integrity without a change of the function and natural frequency before and after seismic loads.

Seismic Analysis of Absorber Rod in KMRR Reactivity Control Mechanism (다목적연구로 반응도 제어장치의 제어봉에 대한 내진해석)

  • Cho, Yeong-Carp;Yoo, Bong;Kim, Tae-Ryong;Ahn, Kyu-Suk
    • Computational Structural Engineering
    • /
    • v.3 no.3
    • /
    • pp.141-146
    • /
    • 1990
  • This study is on a seismic analysis of absorber rod in KMRR Reactivity Control Mechanism. The model being studied is two coaxial tubes(control absorber rod and flow tube) immersed in the water and partially coupled(overlap) by water gap. The hydrodynamic mass effects by the water in each surrounding conditions are considered in the model. The natural frequencies, stresses and displacements of the system due to Safe Shutdown Earthquake are computed in the cases of in-phase modes and out-of-phase modes of two coaxial tubes. The results show that maximum stresses are well below the allowable limit but the maximum displacements at the ends of both tubes are so much that the absorber rod contacts with the flow tube(or surrounding wall).

  • PDF

Dynamic Characteristics on the CRDM of SMART Reactor (SMART 원자로 제어봉 구동 장치의 동특성해석)

  • Lee, Jang-Won;Cho, Sang-Soon;Kim, Dong-Ok;Park, Jin-Seok;Lee, Won-Jae
    • Transactions of the Korean Society of Mechanical Engineers A
    • /
    • v.34 no.8
    • /
    • pp.1105-1111
    • /
    • 2010
  • The Korea Atomic Energy Research Institutes has been developing the SMART (System integrated Modular Advanced ReacTor), an environment-friendly nuclear reactor for the generation of electricity and to perform desalination. SMART reactors can be exposed to various external and internal loads caused by seismic and coolant flows. The CRDM(control rod drive mechanism), one of structures of the SMART, is a component which is adjusting inserting amount of a control rod, controlling output of reactor power and in an emergency situation, inserting a control rod to stop the reactor. The purpose of this research is performing the analysis of dynamic characteristic to ensure safety and integrity of structure of CRDM. This paper presents two FE-models, 3-D solid model and simplified Beam model of the CRDM in the coolant, and then compared the results of the dynamic characteristic about the two FE-models using a commercial Finite Element tool, ABAQUS CAE V6.8 and ANSYS V12. Beam 4 and beam 188 of simplified-model were also compared each other. And simplified model is updated for accuracy compare to 3-D solid.

Development of A Computer Program for Drop Time and Impact Velocity of the Rod Cluster Control Assembly (제어봉집합체의 낙하시간과 충격속도 계산을 위한 프로그램 개발)

  • Park, Ki-Seong;Kim, Il-Kon
    • Nuclear Engineering and Technology
    • /
    • v.26 no.2
    • /
    • pp.197-204
    • /
    • 1994
  • In a PWR rod cluster control assembly(RCCA) for shutdown is released upon action of control rod drive mechanism and falls down through the guide thimble by its weight. Drop time and impact velocity of the RCCA are two key parameters with respect to reactivity insertion time and the mechanical integrity of fuel assembly. Therefore, the precise control of drop time and impact velocity is prerequisite to modifying the existing design features of the RCCA and guide thimble or newly designing them. During its falling down into the core, the RCCA is retarded by various forces acting on it such as fluid resistance caused by the RCCA movement, buoyance and mechanical friction caused by contacting inner surface of the guide thimble, etc. However, complicated coupling of the various forces makes it difficult to derive an analytical dynamic equation for the drop time and impact velocity. This paper deals with the development of a computer program containing an analytical dynamic equation applicable to the Korean Fuel Assembly(KOFA). The computer program is benchmarked with an available single control rod drop tests. Since the predicted values are in good agreement with the test results, the computer program developed in this paper can be employed to modify the exiting design features of the RCCA and guide thimble and to develope their new design features for advanced nuclear reactors.

  • PDF

Analysis of Loss of HVAC for Nuclear Power Plant (원전의 공기조화설비(HVAC) 상실사고 분석방법)

  • Song, Dong-Soo
    • Journal of Energy Engineering
    • /
    • v.23 no.1
    • /
    • pp.90-94
    • /
    • 2014
  • Environmental qualification (EQ) for safety-related equipment is required to ensure that those equipment will perform their required function even under the harsh environment conditions arising from design basis accident in the nuclear power plant. As a part of EQ program, the room temperature analysis in case of a loss of Heating, Ventilation, and Air Conditioning(HVAC) system was carried out to ensure the operability of the safety-related equipment of a nuclear power plant randomly chosen among the Korean nuclear power plants. In this paper, this analysis was performed in the conservative perspective using GOTHIC code. The room temperature analysis includes selecting the rooms in which the safety related equipment are located but not supported by safety related HVAC and determining the temperature of the selected rooms. Target rooms for the analysis consist of W229/W237 (Aux. feedwater pump room), W232 (Aux. feedwater tank room) and W230 (Equipment passageway). The results showed the temperature range from $43^{\circ}C$ to $83^{\circ}C$, in 72 hours after a loss of HVAC. Those values are far below of generic EQ temperature($171^{\circ}C$). Therefore, it is satisfied with EQ requirement of temperature limits on safety related equipment.