• 제목/요약/키워드: Pressure water reactors

검색결과 105건 처리시간 0.025초

HORIZON EXPANSION OF THERMAL-HYDRAULIC ACTIVITIES INTO HTGR SAFETY ANALYSIS INCLUDING GAS-TURBINE CYCLE AND HYDROGEN PLANT

  • No, Hee-Cheon;Yoon, Ho-Joon;Kim, Seung-Jun;Lee, Byeng-Jin;Kim, Ji-Hwang;Kim, Hyeun-Min;Lim, Hong-Sik
    • Nuclear Engineering and Technology
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    • 제41권7호
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    • pp.875-884
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    • 2009
  • We present three nuclear/hydrogen-related R&D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA, with which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed a tool to predict the characteristics of HTGR helium turbines based on the throughflow calculation with a Newton-Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The dynamic equations for the distillation column of HI process are described with 4 material components involved in the HI process: H2O, HI, I2, H2. For the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. The improved Neumann model predicted a total pressure with 8.6% maximum relative deviation from the data and 2.5% mean relative deviation, and liquid-liquid-separation with 9.52% maximum relative deviation from the data.

RADIOLOGICAL CHARACTERISTICS OF DECOMMISSIONING WASTE FROM A CANDU REACTOR

  • Cho, Dong-Keun;Choi, Heui-Joo;Ahmed, Rizwan;Heo, Gyun-Young
    • Nuclear Engineering and Technology
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    • 제43권6호
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    • pp.583-592
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    • 2011
  • The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

사이펀 차단기 시뮬레이션 프로그램의 개발 및 활용 (Development and Application of Siphon Breaker Simulation Program)

  • 이권영;김완수
    • 한국산학기술학회논문지
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    • 제17권5호
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    • pp.346-353
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    • 2016
  • 일부 연구용 원자로의 설계조건상 사이펀 현상은 배관 파단 사고 시 수조수의 지속적인 방출을 유발할 수 있다. 사이펀 차단기는 이러한 현상을 효과적으로 제한하기 위한 안전장치로, 유체역학적인 특성상 사이펀 차단 현상 해석을 위해 고려해야 할 변수가 많고 계산이 복잡하다. 이에 사이펀 차단 현상을 쉽게 분석할 수 있는 프로그램을 개발하게 되었다. 윈도우8 운영체제에서 비쥬얼 스튜디오 2012를 이용하여 MFC프로그래밍으로 개발되었으며, 사용자가 쉽게 사용할 수 있도록 GUI형식으로 개발되었다. 개발된 프로그램은 사용자가 입력한 값으로부터 유체역학적 관계식을 통해 3단계의 연산과정을 거쳐 시뮬레이션을 진행한다. 베르누이 방정식으로부터 유속과 유량을 구하여 수위, 언더슈팅, 압력, 손실계수, 그리고 이상 유동과 관계된 값들을 연산한다. 프로그램에 적용된 이상유동 해석모델은 Chisholm 모델이며, 실제와 유사하게 시뮬레이션이 가능함을 확인하였다. 시뮬레이션 결과는 그래프를 통해 나타나기 때문에 사용자는 전체적인 차단 현상을 쉽게 파악하는 것이 가능하며, 시뮬레이션 데이터의 저장 또한 가능하다. 따라서 사용자는 사이펀 차단기 시뮬레이션 프로그램의 사용을 통해 사이펀 차단 현상을 쉽게 확인할 수 있으며, 사이펀 차단기의 실제 설계에도 이용할 수 있을 것으로 기대된다.

상대유량 모델내의 기포분포계수 측정에 관한 연구 (Measurements of Void Concentration Parameters in the Drift-Flux Model)

  • 윤병조;박군철;정창현
    • Nuclear Engineering and Technology
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    • 제25권1호
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    • pp.91-101
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    • 1993
  • 가압경수로형 원자로의 정상 비정상 운전시의 열수력학적 거동을 예측하기 위해서는 원자로내기포계수의 분포를 정확히 계산하는 것이 필수적이다. 이러한 기포계수의 정확한 예측을 위하여 많은 모델들이 제시되었다. 이중 drift-flux모델은 그 계산의 정확성과 간결성에 의하여 널리 사용되고 있다. 이러한 drift-flux 모델을 사용하여 보다 더 정확한 기포계수를 예측하기 위해서는 각 상간의 슬립률과 flow regime 에 따른 기포의 운동의 변화가 정확히 고려되어야 한다. Drift-flux 모델에서는 이러한 두 가지 요소가 drift-flux parameter인 $C_{o}$ 와 (equation omitted), 에서 고려된다. 본 연구에서는 이러한 $C_{o}$ 의 실험적 결정을 위하여 원자로 노심을 모사한 4개의 전열봉이 있는 비등이 발생하는 수직사각 유로를 구성하였으며, 완성된 유로내에서 기포계수의 분포 및 기포속도의 분포를 측정하였다. 국부적 기포계수 및 기포속도 분포의 측정에 사용된 방법은 이중탐침법이며 측정이 이루어진 유로내의 유동 상태는 유속이 비교적 느린 low flow rate condition이며 유로내 압력은 3기압 이하이다. 본 실험에서는 액상의 속도는 측정되지 않았으며, 따라서 $C_{o}$ 의 계산을 위하여 (equation omitted)의 실험 상관관계식을 사용하여 유로내 평균 기포계수의 함수로 나타내었다.

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Research of Diffusion Bonding of Tungsten/Copper and Their Properties under High Heat Flux

  • Li, Jun;Yang, Jianfeng
    • 한국재료학회:학술대회논문집
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    • 한국재료학회 2011년도 춘계학술발표대회
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    • pp.14-14
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    • 2011
  • W (tungsten)-alloys will be the most promising plasma facing armor materials in highly loaded plasma interactive components of the next step fusion reactors due to its high melting point, high sputtering resistance and low deuterium/tritium retention. The bonding technology of tungsten to Cu alloy was one of the key issues. In this paper, W/CuCrZr diffusion bonding has been performed successfully by inserting pure metal interlay. The joint microstructure, interfacial elements migration and phase composition were analyzed by SEM, EDS, XRD, and the joint shear strength and micro-hardness were investigated. The mock-ups were fabricated successfully with diffusion bonding and the cladding technology respectively, and the high heat flux test and thermal fatigue test were carried out under actively cooling condition. When Ni foil was used for the bonding of tungsten to CuCrZr, two reaction layers, Ni4W and Ni(W) layer, appeared between the tungsten and Ni interlayer with the optimized condition. Even though Ni4W is hard and brittle, and the strength of the joint was oppositely increased (217 MPa) due primarily to extremely small thicknesses (2~3 ${\mu}m$). When Ti foil was selected as the interlayer, the Ti foil diffused quickly with Cu and was transformed into liquid phase at $1,000^{\circ}C$. Almost all of the liquid was extruded out of the interface zone under bonding pressure, and an extremely thin residual layer (1~2 ${\mu}m$) of the liquid phase was retained between the tungsten and CuCrZr, which shear strength exceeded 160 MPa. When Ni/Ti/Ni multiple interlayers were used for bonding of tungsten to CuCrZr, a large number of intermetallic compound ($Ni_4W/NiTi_2/NiTi/Ni_3T$) were formed for the interdiffusion among W, Ni and Ti. Therefore, the shear strength of the joint was low and just about 85 MPa. The residual stresses in the clad samples with flat, arc, rectangle and trapezoid interface were estimated by Finite Element Analysis. The simulation results show that the flat clad sample was subjected maximum residual stress at the edge of the interface, which could be cracked at the edge and propagated along the interface. As for the rectangle and trapezoid interface, the residual stresses of the interface were lower than that of the flat interface, and the interface of the arc clad sample have lowest residual stress and all of the residual stress with arc interface were divided into different grooved zones, so the probabilities of cracking and propagation were lower than other interfaces. The residual stresses of the mock-ups under high heat flux of 10 $MW/m^2$ were estimated by Finite Element Analysis. The tungsten of the flat interfaces was subjected to tensile stresses (positive $S_x$), and the CuCrZr was subjected to compressive stresses (negative $S_x$). If the interface have a little microcrack, the tungsten of joint was more liable to propagate than the CuCrZr due to the brittle of the tungsten. However, when the flat interface was substituted by arc interfaces, the periodical residual stresses in the joining region were either released or formed a stress field prohibiting the growth or nucleation of the interfacial cracks. Thermal fatigue tests were performed on the mock-ups of flat and arc interface under the heat flux of 10 $MW/m^2$ with the cooling water velocity of 10 m/s. After thermal cycle experiments, a large number of microcracks appeared at the tungsten substrate due to large radial tensile stress on the flat mock-up. The defects would largely affect the heat transfer capability and the structure reliability of the mock-up. As for the arc mock-up, even though some microcracks were found at the interface of the regions, all microcracks with arc interface were divided into different arc-grooved zones, so the propagation of microcracks is difficult.

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