• 제목/요약/키워드: loss of coolant accident

검색결과 201건 처리시간 0.023초

냉각속도가 지르칼로이-4 피복관의 취성에 미치는 영향 (Effect of Cooling Rate on the Behavior of the Embrittlement in Zircaloy-4 Cladding)

  • 김준환;이명호;최병권;정용환
    • 열처리공학회지
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    • 제18권2호
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    • pp.112-118
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    • 2005

  • Study was focused on the effect of the cooling rate on the embrittlement behavior of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment. Claddings were oxidized at given temperature and given time followed by various water quenching in the range of $0.6^{\circ}C$ and $100^{\circ}C$ per second. Cladding failed after water quenching above the threshold oxidation. Threshold oxidation was decreased as the cooling rate increased, which is due to the matensite structure formed during fast cooling rate.

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The corrosion of aluminium alloy and release of intermetallic particles in nuclear reactor emergency core coolant: Implications for clogging of sump strainers

  • Huang, Junlin;Lister, Derek;Uchida, Shunsuke;Liu, Lihui
    • Nuclear Engineering and Technology
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    • 제51권5호
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    • pp.1345-1354
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    • 2019

  • Clogging of sump strainers that filter the recirculation water in containment after a loss-of-coolant accident (LOCA) seriously impedes the continued cooling of nuclear reactor cores. In experiments examining the corrosion of aluminium alloy 6061, a common material in containment equipment, in borated solutions simulating the water chemistry of sump water after a LOCA, we found that Fe-bearing intermetallic particles, which were initially buried in the Al matrix, were progressively exposed as corrosion continued. Their cathodic nature $vis-{\grave{a}}-vis$ the Al matrix provoked continuous trenching around them until they were finally released into the test solution. Such particles released from Al alloy components in a reactor containment after a LOCA will be transported to the sump entrance with the recirculation flow and trapped by the debris bed that typically forms on the strainer surface, potentially aggravating strainer clogging. These Fe-bearing intermetallic particles, many of which had a rod or thin strip-like geometry, were identified to be mainly the cubic phase ${\alpha}_c-Al(Fe,Mn)Si$ with an average size of about $2.15{\mu}m$; 11.5 g of particles with a volume of about $3.2cm^3$ would be released with the dissolution of every 1 kg 6061 aluminium alloy.

  • https://doi.org/10.1016/j.net.2019.02.012 인용 PDF KSCI

Mechanical analysis for prestressed concrete containment vessels under loss of coolant accident

  • Zhou, Zhen;Wu, Chang;Meng, Shao-ping;Wu, Jing
    • Computers and Concrete
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    • 제14권2호
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    • pp.127-143
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    • 2014

  • LOCA (Loss Of Coolant Accident) is one of the most important utmost accidents for Prestressed Concrete Containment Vessel (PCCV) due to its coupled effect of high temperature and inner pressure. In this paper, heat conduction analysis is used to obtain the LOCA temperature distribution of PCCV. Then the elastic internal force of PCCV under LOCA temperature is analyzed by using both simplified theoretical method and FEM (finite element methods) method. Considering the coupled effect of LOCA temperature, a nonlinear elasto-plasitic analysis is conducted for PCCV under utmost internal pressure considering three failure criteria. Results show that the LOCA temperature distribution is strongly nonlinear along the shell thickness at the early time; the moment result of simplified analysis is well coincident with the one of numerical analysis at weak constraint area; while in the strong constrained area, the value of moments and membrane forces fluctuate dramatically; the simplified and numerical analysis both show that the maximum moment occurs at 6hrs after LOCA.; the strain of PCCV under LOCA temperature is larger than the one of no temperature under elasto-plastic analysis; the LOCA temperature of 6hrs has the greatest influence on the ultimate bearing capacity with 8.43% decrease for failure criteria 1 and 2.65% decrease for failure criteria 3.

  • https://doi.org/10.12989/cac.2014.14.2.127 인용 KSCI

가압열충격 사고에 대한 원자로 용기의 최대 허용 기준무연성천이온도 (Maximum Allowable $RT_{NDT}$ of Nuclear Reactor Vessel for Pressurized Thermal Shock Accident)

  • 정명조;박윤원;송선호
    • 전산구조공학
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    • 제11권1호
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    • pp.153-160
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    • 1998

  • 본 연구에서는 가압열충격 사고로 소형 냉각재 상실사고를 가정하여 냉각재의 온도와 압력의 이력으로 부터 용기 벽의 온도분포를 구하고, 이로 부터 열응력과 압응력을 해석적으로 구하였다. 또 균열 선단에서의 응력강도계수와 파괴인성치를 ASME코드의 방법을 이용하여 구하였고, 이들을 시간에 따라 비교하여 균열의 진전여부를 평가하였다. 원자로 용기 벽에 존재하는 여러 형태의 균열이 견딜 수 있는 최대 기준무연성천이온도를 결정하였으며 평가 결과에 대하여 고찰하였다.

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