• Title/Summary/Keyword: Containment safety

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Assessment of the MELCOR 1.8.6 condensation heat transfer model under the presence of noncondensable gases (중대사고 해석코드 MELCOR 1.8.6의 비응축성기체 존재 시 응축열전달 모델 평가)

  • Yoo, Ji Min;Lee, Dong Hun;Yun, Byong Jo;Jeong, Jae Jun
    • Journal of Energy Engineering
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    • v.25 no.2
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    • pp.1-20
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    • 2016
  • A condensation heat transfer model is very important for the safety analysis of nuclear power plants. Especially, condensation under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NCGs in the vapor largely reduces the condensation rate. In this study, the condensation heat transfer model of the severe accident analysis code MELCOR 1.8.6 has been assessed using a set of condensation experiments performed under the thermal-hydraulic conditions similar to those inside a containment during design-basis accidents or severe accidents. Experiment conditions are categorized into 4 types according to the shape of the condensation surface: vertical flat plates, outer surface of vertical pipes, inner surface of vertical pipes, the inner surface of horizontal pipes. The results of the calculations show that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on inner surface of vertical pipes.

Experimental Study on Fire-Resistant Characteristics of Bi-Directionally Prestressed Concrete Panel under RABT Fire Scenario (RABT 화재시나리오를 적용한 이방향 프리스트레스트 콘크리트 패널부재의 내화특성에 관한 실험적 연구)

  • Yi, Na-Hyun;Lee, Sang-Won;Kim, Jang-Ho Jay
    • Journal of the Korea Concrete Institute
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    • v.24 no.6
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    • pp.695-703
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    • 2012
  • Recently, major infrastructure such as bridges, tunnels, PCCVs (Prestressed Concrete Containment Vessel), and gas tanks are Prestressed Concrete (PSC) structure types, which improve their safety by using confining effect from prestressing. Generally, concrete is known to be an outstanding fire resistant construction material. Because of this reason, researches related to extreme fire loaded PSC member behaviors are not often conducted even though PSC behavior under extreme fire loading is significantly different than that of ordinary reinforced concrete (RC) behavior. Therefore, in this study, RABT fire loading tests were performed on bi-directionally prestressed concrete panels with $1000{\times}1400{\times}300mm$ dimensions. The prestressed specimens were applied with 430 kN prestressing (PS) force using unbonded PS thread bars. Also, residual strength structural tests of fire tested PSC and ordinary RC structures were performed for comparison. The study results showed that PSC behavior under fire loading is significantly different than that of RC behavior.

Numerical Evaluation of Debris Transport During LOCA Blow-Down Phase of OPR1000 Nuclear Power Plant (CFD 를 이용한 OPR1000 원자력발전소 파단방출이동에 대한 수치해석적 평가)

  • Choi, Kyung-Sik;Park, Jong-Pil;Jeong, Ji-Hwan;Kim, Won-Tae
    • Transactions of the Korean Society of Mechanical Engineers B
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    • v.35 no.3
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    • pp.255-262
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    • 2011
  • In a loss-of-coolantaccident, considerable debris may be generated and transported to the recirculation sump. The accumulation of debris will reduce the netpositivesuctionhead and threaten the safety of thenuclear power plant. Both NEI 04-07 and USNRC SER suggesteda CFD methodology. However, additional investigation is needed to consider the unique characteristics of nuclear power plants. The transport of the generated debris is strongly influenced by the break location and the plant characteristics, including the configuration.In this paper, a CFD methodology for blow-down transport evaluation is proposed and applied to an OPR1000 nuclear power plant. The results show that the percentage of small debris transported to the upper containment is 32%, which is 7% larger than the valuegiven in the NEI 04-07 baseline analysis. This result may be used as a point of reference in future analytical studies.

Analysis Evaluation of Impact Behavior of 270,000kL LNG Storage Outer Tank from Prestress Force Loss (프리스트레스 손실량을 고려한 270,000kL급 LNG 저장탱크 외조의 비산물체 속도에 따른 충돌 거동 해석)

  • Lee, Sang-Won;Jun, Ha-Young;Kim, Jang-Ho Jay;Kim, Jun-Hwi;Lee, Kang-Won
    • Journal of the Korean Institute of Gas
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    • v.18 no.1
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    • pp.31-40
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    • 2014
  • LNG storage outer tank is a vertically and horizontally prestressed concrete wall structure. Therefore, when the storage tanks become larger, prestressing tendons become longer and eventually the prestressing loss becomes larger. Also, recently, bomb terrors and accidental crashes have occurred frequently on important infrastructures. Therefore, LNG storage tanks are also exposed to these dangerous scenarios, where they need to be evaluated and protected from these threats. Therefore, in this study, the behavior of 270,000 kL LNG storage outer tank impacted by a flying object is evaluated using implicit FEM code, LS-DYNA. In the analysis, the prestress loss due to the increased length of prestressing tendons from enlargement of outer tank is considered. A comparison study between the LNG tanks with and without prestress loss is performed to investigate the impact behavior and the effect of prestressing force change on the safety and serviceability prestressed concrete containment.

Static and Dynamic Analysis of Reinforced Concrete Axisymmetric Shell on the Elastic Foundation -Effect of Steel on the Dynamic Response- (탄성지반상에 놓인 철근 콘크리트 축대칭 쉘의 정적 및 동적 해석(IV) -축대칭 쉘의 동적 응답에 대한 철근의 영향을 중심으로-)

  • 조진구
    • Magazine of the Korean Society of Agricultural Engineers
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    • v.39 no.4
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    • pp.106-113
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    • 1997
  • Dynamic loading of structures often causes excursions of stresses well into the inelastic range, and the influence of the geometric changes on the dynamic response is also significant in many cases. Therefore, both material and geometric nonlinearity effects should be considered in case that a dynamic load acts on the structure. A structure in a nuclear power plant is a structure of importance which puts emphasis on safety. A nuclear container is a pressure vessel subject to internal pressure and this structure is constructed by a reinforced concrete or a pre-stressed concrete. In this study, the material nonlinearity effect on the dynamic response is formulated by the elasto-viscoplastic model highly corresponding to the real behavior of the material. Also, the geometrically nonlinear behavior is taken into account using a total Lagrangian coordinate system, and the equilibrium equation of motion is numerically solved by a central difference scheme. The constitutive relation of concrete is modeled according to a Drucker-Prager yield criterion in compression. The reinforcing bars are modeled by a smeared layer at the location of reinforcements, and the steel layer model under Von Mises yield criteria is adopted to represent an elastic-plastic behavior. To investigate the dynamic response of a nuclear reinforced concrete containment structure, the steel-ratios of 0, 3, 5 and 10 percent, are considered. The results obtained from the analysis of an example were summarized as follows 1. As the steel-ratio increases, the amplitude and the period of the vertical displacements in apex of dome decreased. The Dynamic Magnification Factor(DMF) was some larger than that of the structure without steel. However, the regular trend was not found in the values of DMF. 2. The dynamic response of the vertical displacement and the radial displacement in the dome-wall junction were shown that the period of displacement in initial step decreased with the steel-ratio increases. Especially, the effect of the steel on the dynamic response of radial displacement disapeared almost. The values of DMF were 1.94, 2.5, 2.62 and 2.66, and the values increased with the steel-ratio. 3. The characteristics of the dynamic response of radial displacement in the mid-wall were similar to that of dome-wall junction. The values of DMF were 1.91, 2.11, 2.13 and 2.18, and the values increased with the steel-ratio. 4. The amplitude and the period of the hoop-stresses in the dome, the dome-wall junction, and the mid-wall were shown the decreased trend with the steel-ratio. The values of DMF were some larger than those of the structure without steel. However, the regular trend was not found in the values of DMF.

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Development of a Raman Lidar System for Remote Monitoring of Hydrogen Gas (수소 가스 원격 모니터링을 위한 라만 라이다 시스템 개발)

  • Choi, In Young;Baik, Sung Hoon;Park, Nak Gyu;Kang, Hee Young;Kim, Jin Ho;Lee, Na Jong
    • Korean Journal of Optics and Photonics
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    • v.28 no.4
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    • pp.166-171
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    • 2017
  • Hydrogen gas is a green energy sources because it features no emission of pollutants during combustion. But hydrogen gas is very dangerous, being flammable and very explosive. Hydrogen gas detection is very important for the safety of a nuclear power plant. Hydrogen gas is generated by oxidation of nuclear fuel cladding during a critical accident, and leads to serious secondary damage in the containment building. This paper discusses the development of a Raman lidar system for remote detection and measurement of hydrogen gas. A small, portable Raman lidar system was designed, and a measurement algorithm was developed to quantitatively measure hydrogen gas concentration. To verify the capability of measuring hydrogen gas with the developed Raman lidar system, experiments were carried out under daytime outdoor conditions by using a gas chamber that can adjust the hydrogen gas density. As results, our Raman lidar system is able to measure a minimum density of 0.67 vol. % hydrogen gas at a distance of 20 m.

Determination of Design Basis for a Storage System for Spent Fuel in Korea (국내 사용후핵연료 저장시스템의 설계기준 설정 인자 고찰)

  • Yoon, Jeong-Hyoun;Lee, Eun-Yong;Woo, Sang-In;Kim, Tae-Man
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.9 no.2
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    • pp.113-119
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    • 2011
  • Safe operation and maintenance of engineered dry storage systems for spent fuel from nuclear power plants basically depends on adequately adopted design requirements. The most important design target of the system are those which provide the necessary assurances that spent fuel can be received, handled, stored and retrieved without undue risk to health and safety of workers and the public. To achieve these objectives, the design of the system incorporates features to remove spent fuel residual heat, to provide for radiation protection, and to maintain containment over the lifespan of the system as specified in the design specifications. The features also provide for all possible anticipated operational occurrences and design basis events in accordance with the design basis as guided by the designated regulations. The general performance requirements of a projected storage system are introduced in this paper. The storage system is designed to store fuel assemblies in associated with designated regulatory requirements. Small increases/decreases in maximum burnup can be adjusted with cooling time. These variations are compensated for by a corresponding small site-specific increase/decrease in the design basis-cooling period, as long as the maximum heat load and radioactivity of loaded fuel assemblies are met. Generic design basis events considered for the storage system are summarized. Shielding and radiological requirements along with mechanical and structural are derived in this study.

A Study on Physicochemical Properties of Epoxy Coatings for Liner Plate in Nuclear Power Plant (원자력발전소 격납건물 철재면 에폭시 도장시편의 물리화학적 특성 평가)

  • Lee, Jae-Rock;Seo, Min-Kang;Lee, Sang-Kook;Lee, Chul-Woo;Park, Soo-Jin
    • Applied Chemistry for Engineering
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    • v.16 no.6
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    • pp.809-814
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    • 2005
  • In this work, the thermal properties of epoxy coating system on the liner plate in the containment structure of nuclear power plants had been examined by irradiation and design basis accident (DBA) conditions. The effect of immersion in hot water on adhesion strength of the coating system had been also studied. The glass transition temperature ($T_g$) and thermal stability of ET-5290/carbon steel A 32 epoxy coating systems were measured by DSC and TGA analyses, respectively. Contact angle measurements were used to determine the effect of immersion on the surface energetics of epoxy coating system, with a viewpoint of surface free energy. Adhesion tests were also executed to evaluate the adhesion strength at interfaces between carbon steel plate and epoxy resins. As a result, it was found that the irradiation led to an improvement of internal crosslinked structure in cured epoxy systems, resulting in significantly increasing the thermal stability, as well as the $T_g$. Also, the immersion in hot water made a role in the post-curing of epoxy resins and increased the mechanical interlocking of the network system, resulting in increasing the adhesion strength of the epoxy coating system.