• Title/Summary/Keyword: Reactor safety

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Design of a Dynamic Absorber for the Large-Size Pressure Vessel of the Petrochemical Plant (석유화학 플랜트의 대형 압력용기에 대한 동흡진기의 설계)

  • Kim, Min-Chul;Lee, Boo-Youn;Kim, Won-Jin
    • Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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    • 2005.05a
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    • pp.743-749
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    • 2005
  • In this work, two dynamic absorbers are introduced and designed to reduce the vibration of the large-size pressure vessel of a reactor for a petrochemical plant. The vibration modes and harmonic responses of the vessel are firstly analyzed by the finite element method. On the basis of the analyzed results, two dynamic absorbers are designed by a simple design theory. Furthermore, an optimization process is executed and an optimal design of the dynamic absorber is obtained to improve performance and structural safety of the vessel. As a result, the maximum displacement and stress of the vessel is decreased about 85% and 65% respectively, the design criteria being satisfied.

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A Scoping Analysis of Venting Capability During Loss of RHRS Events

  • Lee, Cheol-Sin;Han, Kee-Soo;Park, Chul-Jin;Kim, Hee-Cheol
    • Proceedings of the Korean Nuclear Society Conference
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    • 1996.05b
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    • pp.657-662
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    • 1996
  • Venting capability to prevent excess pressurization caused by loss of Residual Heat Removal System (RHRS) during mid-loop operation hat been evaluated analytically and the peak Reactor Coolant System (RCS) pressure was compared with the results of the MIDLOOP computer code. Even though analytical method if relatively simple, the results are in a good agreement with those of the computer code. For both methods, the peak pressures have not, exceeded the nozzle dam design pressure, if the vent paths such as pressurizer safety valves or a pressurizer manway are available in a closed RCS configuration with the nozzle dam installed.

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RTD Bypass Line Elimination at Kori Nuclear Power Plant 3&4

  • Yoon, Duk-Joo;Lee, Chang-Sup;Jun, Hwang-Yong;Lee, Jae-Yong;Song, Dong-Soo
    • Proceedings of the Korean Nuclear Society Conference
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    • 1997.05a
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    • pp.213-218
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    • 1997
  • The RTD Bypass Loops at Kori Unit 3&4 will be removed and a new system will be designed and will be installed to replace it. The replacement system provides equal or better performance and eliminates some Persistent problems. The Resistance Temperature Detector (RTD) bypass line is eliminated to reduce the radiation exposure to operators and workers. After the elimination, the resistance temperature detectors are installed in scoop of the reactor coolant piping to detect a representative temperature. This study includes safety evaluation, RTD response time Analysis, Uncertainty Analysis, LOCA evaluation and Structural Analysis.

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Analysis of Common Cause Failure Using Two-Step Expectation and Maximization Algorithm (2단계 EM 알고리즘을 이용한 공통원인 고장 분석)

  • Baek Jang Hyun;Seo Jae Young;Na Man Gyun
    • Journal of the Korean Operations Research and Management Science Society
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    • v.30 no.2
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    • pp.63-71
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    • 2005
  • In the field of nuclear reactor safety study, common cause failures (CCFs) became significant contributors to system failure probability and core damage frequency in most Probabilistic risk assessments. However, it is hard to estimate the reliability of such a system, because of the dependency of components caused by CCFs. In order to analyze the system, we propose an analytic method that can find the parameters with lack of raw data. This study adopts the shock model in which the failure probability increases as the shock is cumulated. We use two-step Expectation and Maximization (EM) algorithm to find the unknown parameters. In order to verify the analysis result, we perform the simulation under same environment. This approach might be helpful to build the defensive strategy for the CCFs.

Response Instrumentation Test Acceptance Criteria for APR1400 RVI CVAP (APR1400 원자로내부구조물 종합진동평가 응답측정시험 허용기준)

  • Ko, Do-Young;Kim, Kyu-Hyung;Kim, Sung-Hwan
    • Transactions of the Korean Society for Noise and Vibration Engineering
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    • v.21 no.11
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    • pp.1036-1042
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    • 2011
  • APR1400 RVI CVAP using the non-prototype category II is being conducted to verify integrity of the RVI design and to secure the CVAP technology. The measurement programs are to confirm vibration analysis results for reactor internals during pre-operational and initial startup testing and to determine the safety margin. One of the important basis for the measurement programs is test acceptance criteria. Therefore, this paper is on establishment of response instrumentation test acceptance criteria for APR1400 RVI CVAP. The established acceptance criteria show that the stress criteria of APR1400 RVI are more conservative values than those of the valid prototype plant(Palo Verde unit 1) and, the displacement criteria of the inner barrel assembly and the upper guide structure were established to 0.03 in and 0.01 in, respectively.

A Study on the Analysis of Failures Related to Emergency Diesel Generators in Overseas Nuclear Power Plants (원전용 비상디젤발전기 국외 손상사례 분석에 관한 연구)

  • Chang, Jung-Hwan;Kim, Jin-Sung;Chung, Hae-Dong;Cho, Kwon-Hae
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.5 no.1
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    • pp.32-37
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    • 2009
  • The emergency diesel generator (EDG) in a nuclear power plant (NPP) shall start within 10 secondss and supply electrical power to engineered safety features within one minute and less if a loss of offsite power (LOOP), A design-basis event, or their combination occur. Each NPP has an EDG set consisting of two diesel generators for redundancy. In addition to the EDG set, an alternate Alternating Current Diesel Generator (AAC DG) is installed and shared by several units to cope with a station black out (SBO), i.e., loss of the offsite power concurrent with reactor trip and unavailability of the EDG set. The objective of this study is to analyze the failure data of emergency diesel generators reported in overseas nuclear power plants.

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Design and Manufacturing of Control Rod Control System for Nuclear Power System (원전 적용을 위한 제어봉 구동장치 제어시스템 설계 및 제작)

  • Lee, J.M.;Kim, C.K.;Kim, S.J.;Cheon, J.M.;Shin, J.R.;Kweon, S.M.;Nam, J.H.
    • Proceedings of the KIEE Conference
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    • 2004.07d
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    • pp.2298-2300
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    • 2004
  • This paper does with the design. implementation, and test of a CRCS for nuclear power plants. Although CRCS is still classified into non-safety class, much attention on its reliability issue has been given so far because of its importance for the stable operation of the reactor in the plant. In terms of technical aspects, our system adopts a full-duplex configuration to enhance reliability in contrast to the existing systems that are all simplex.

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Numerical Analysis of Thermal Stratification and Turbulence Penetration into Leaking Flow in a Circular Branch Piping (원형 T분기배관 내 누설유동의 열성층화와 난류침투에 관한 전산해석적 연구)

  • Han, Seong-Min;Choi, Young-Don
    • Proceedings of the KSME Conference
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    • 2003.04a
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    • pp.1833-1838
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    • 2003
  • In the nuclear power plant, emergency core coolant system(ECCS) is furnished at reactor coolant system(RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can be occurred due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack(TFC) accident. In the present study, when the turbulence penetration occurs in the branch piping, the maximum temperature differences of fluid at the pipe cross-sections of the T-branch with thermal stratification are examine

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A Study of the Evaporation Heat Transfer in Advanced Reactor Containment

  • Y. M. Kang;Park, G. C.
    • Nuclear Engineering and Technology
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    • v.29 no.4
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    • pp.291-298
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    • 1997
  • In advanced nuclear reactors, the passive containment cooling has been suggested to enhance the safety. The passive cooling has two mechanisms, air natural convection and oater cooling with evaporation. To confirm the coolability of PCCS, many works have been performed experimentally and numerically. In this study, the water cooling test was performed to obtain the evaporative heat transfer coefficients in a scaled don segment type PCCS facility which have same configuration with AP600 prototype containment. Air-steam mixture temperature and velocity, relative humidity and well heat flux are measured. The local steam mass flow rates through the vertical plate part of the facility are calculated from the measured data to obtain evaporative heat transfer coefficients. The measured evaporative heat transfer coefficients are compared with an analytical model which use a mass transfer coefficients. From the comparison, the predicted coefficients show good agreement with experimental data however, some discrepancies exist when the effect of wave motion is not considered. Finally, a new correlation on evaporative heat transfer coefficients are developed using the experimental values.

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Axial BP Zoning for the Soluble Boron Free Operation in Medium-Sized PWR

  • Kim, Jong-Chae;Kim, Myung-Hyun
    • Proceedings of the Korean Nuclear Society Conference
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    • 1996.11a
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    • pp.59-64
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    • 1996
  • Feasibility of soluble boron free operation for the medium-sized commercial reactors was investigated. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with gadolinia burnable poison-high Gd enrichment and axial poison zoning. CASMO and NECTA-C code system checked axial offset and peaking factors as fuels burned up. A core with complex axial burnable poison zoning satisfied design goals - small excess reactivity for 18 month cycle. Therefore, critical bank positioning for three control rod banks was sought with ease. A.O. value and Fq value were kept within the safety limit.

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