• 제목/요약/키워드: integrity assessment

검색결과 554건 처리시간 0.03초

VIBRATION AND STRESS ANALYSIS OF A UGS ASSEMBLY FOR THE APR1400 RVI CVAP

  • Ko, Do-Young;Kim, Kyu-Hyung
    • Nuclear Engineering and Technology
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    • 제44권7호
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    • pp.817-824
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    • 2012
  • The most important component of a nuclear power plant is its nuclear reactor. Studies on the integrity of reactors have become an important part regarding the safety of a nuclear power plant. The US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) to be used to verify the structural integrity of the Reactor Vessel Internals (RVI) for flow-induced vibration prior to commercial operation. However, there are few published studies related to the RVI CVAP. We classified the Advanced Power Reactor 1400 (APR1400) RVI CVAP as a non-prototype category-2 reactor as part of an independent validation of its design. The aim of this paper is to present the results of structural response analyses of the Upper Guide Structure (UGS) assembly of the APR1400 reactor. These results show that the UGS and the Inner Barrel Assembly (IBA) meet the specified integrity levels of the design acceptance criteria. The vibration and stress analysis results in this paper will be used as basic information to select measurement locations of the vibration and stress for the APR1400 RVI CVAP.

원자로 내부구조물의 유체흐름에 의한 진동 - 해석 및 실험 (Flow Induced Vibration of Reactor Internals Structure : Analysis and Experiment)

  • 이희남;최순;김태형;황종근;김정규
    • 한국소음진동공학회:학술대회논문집
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    • 한국소음진동공학회 1995년도 추계학술대회논문집; 한국종합전시장, 24 Nov. 1995
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    • pp.201-207
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    • 1995
  • A series of vibration assessment programs has been performed for Yonggwang Nuclear Power Plant Unit 4 (YGN 4) in order to verify the structural integrity of the reactor internals for flow induced vibration prior to its commercial operation. The structural analysis was done to provide the basis for measurement and the theoretical evidence for the structural integrity of the reactor internals. The actual flow induced hydraulic loads and reactor internals vibration response data were measured and recorded during pre-core hot functional testing of the plant. Then, the measured data have been reduced and analyzed, and compared with the analysis results such as the frequency contents, stresses, strains and displacements. It is concluded that the structural analysis methodology performed for vibration response of the reactor internals due to the flow induced vibration is appropriately conservative, and also that the structural integrity of YGN 4 reactor internals to flow induced vibration is acceptable for long term operation.

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유전자 알고리즘을 이용한 CANDU 압력관의 확률론적 손상역학 평가 (Probabilistic Damage Mechanics Assessment of CANDU Pressure Tube using Genetic Algorithm)

  • 고한옥;장윤석;최재붕;김영진;김홍기;최영환
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2008년도 추계학술대회A
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    • pp.192-192
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    • 2008
  • As the lifetime of nuclear power plants (NPPs) reaches design life, the probability for fatal accidents increases. Most of accidents are known to be caused by degradation of mechanical components. Pressure tubes are the most important components in CANDU reactor. They are subjected to various aging mechanisms such as delayed hydride cracking (DHC), irradiation and corrosion, etc. Therefore, the integrity of pressure tube is key concern in CANDU reactor. Up to recently, conventional deterministic approaches have been utilized to evaluate the integrity of components. However, there are many uncertainties to prevent a rational evaluation. The objective of this paper is to assess the failure probability of pressure tube in CANDU. To do this, probability fracture mechanics (PFM) analysis based on the Genetic Algorithm (GA) is performed. For the verification of the analysis, a comparison of the PFM analysis using a commercial code and mathematical method is carried out.

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몬테카를로 시뮬레이션을 이용한 확률론적 파괴역학 수법의 적용성 검토 (Application of Probabilistic Fracture Mechanics Technique Using Monte Carlo Simulation)

  • 이준성;곽상록;김영진
    • 한국정밀공학회지
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    • 제18권10호
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    • pp.154-160
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    • 2001
  • For major structural components periodic inspections and integrity assessments are needed for the safety. However, many flaws are undetectable because sampling inspection is carried out during in-service inspection. Probabilistic integrity assessment is applied to take into consideration of uncertainty and variance of input parameters arise due to material properties and undetectable cracks. This paper describes a Probabilistic Fracture Mechanics(PFM) analysis based on the Monte Carlo(MC) algorithms. Taking a number of sampling data of probabilistic variables such as fracture toughness value, crack depth and aspect ratio of an initial surface crack, a MC simulation of failure judgement of samples is performed. for the verification of this analysis, a comparison study of the PFM analysis using a commercial code, mathematical method is carried out and a good agreement was observed between those results.

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INTEGRITY ANALYSIS OF AN UPPER GUIDE STRUCTURE FLANGE

  • LEE, KI-HYOUNG;KANG, SUNG-SIK;JHUNG, MYUNG JO
    • Nuclear Engineering and Technology
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    • 제47권6호
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    • pp.766-775
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    • 2015
  • The integrity assessment of reactor vessel internals should be conducted in the design process to secure the safety of nuclear power plants. Various loads such as self-weight, seismic load, flow-induced load, and preload are applied to the internals. Therefore, the American Society of Mechanical Engineers (ASME) Code, Section III, defines the stress limit for reactor vessel internals. The present study focused on structural response analyses of the upper guide structure upper flange. The distributions of the stress intensity in the flange body were analyzed under various design load cases during normal operation. The allowable stress intensities along the expected sections of stress concentration were derived from the results of the finite element analysis for evaluating the structural integrity of the flange design. Furthermore, seismic analyses of the upper flange were performed to identify dynamic behavior with respect to the seismic and impact input. The mode superposition and full transient methods were used to perform time-history analyses, and the displacement at the lower end of the flange was obtained. The effect of the damping ratio on the response of the flange was also evaluated, and the acceleration was obtained. The results of elastic and seismic analyses in this study will be used as basic information to judge whether a flange design meets the acceptance criteria.

연구용 원자로 내부에 설치되는 이차정지구동장치의 내진낙하성능 (Seismic Drop Performance for Second Shutdown Drive Mechanism Installed in Research Reactor)

  • 김상헌;김경호;선종오;조영갑;김정현;정택형;이관희
    • 한국소음진동공학회논문집
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    • 제26권6_spc호
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    • pp.697-704
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    • 2016
  • The second shutdown drive mechanism (SSDM) that is classified into seismic category I as an active mechanical equipment shall maintain the structural integrity and its designed inherent safety functions during and/or after normal operation, anticipated operational occurrences, accidents and seismic occurrences. Therefore, not only a structural integrity assessment through numerical analyses but also a qualification test by using the prototype SSDM shall be conducted to verify the adequacy of the SSDM design. This paper describes a sort of seismic qualification test of the prototype SSDM to demonstrate that the structural integrity and operability (functionality) of SSDM are maintained during and/or after seismic excitations. From the results, this paper shows that the SSDM satisfies all design requirements without any malfunctions during and after the seismic test.

Development of a structural integrity evaluation program for elevated temperature service according to ASME code

  • Kim, Nak Hyun;Kim, Jong Bum;Kim, Sung Kyun
    • Nuclear Engineering and Technology
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    • 제53권7호
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    • pp.2407-2417
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    • 2021
  • A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

Electrical fire simulation in control room of an AGN reactor

  • Jyung, Jae-Min;Chang, Yoon-Suk
    • Nuclear Engineering and Technology
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    • 제53권2호
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    • pp.466-473
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    • 2021
  • Fire protection is one of important issues to ensure safety and reduce risks of nuclear power plants (NPPs). While robust programs to shut down commercial reactors in any fires have been successfully maintained, the concept and associated regulatory requirements are constantly changing or strengthening by lessons learned from operating experiences and information all over the world. As part of this context, it is necessary not only to establish specific fire hazard assessment methods reflecting the characteristics of research reactors and educational reactors but also to make decisions based on advancement encompassing numerical analyses and experiments. The objectives of this study are to address fire simulation in the control room of an educational reactor and to discuss integrity of digital console in charge of main operation as well as analysis results through comparison. Three electrical fire scenarios were postulated and twenty-four thermal analyses were carried out taking into account two turbulence models, two cable materials and two ventilation conditions. Twelve supplementary thermal analyses and six subsequent structural analyses were also conducted for further examination on the temperature and heat flux of cable and von Mises stress of digital console, respectively. As consequences, effects of each parameter were quantified in detail and future applicability was briefly discussed. On the whole, higher profiles were obtained when Deardorff turbulence model was employed or polyvinyl chloride material and larger ventilation condition were considered. All the maximum values considered in this study met the allowable criteria so that safety action seems available by sustained integrity of the cable linked to digital console within operators' reaction time of 300 s.

APR1400 원자로내부구조물 종합진동평가프로그램 진동 및 응력해석 방법론 검증 (Validation of Vibration and Stress Analysis Methodology for APR1400 Reactor Vessel Internals Comprehensive Vibration Assessment Program)

  • 김규형;고도영;김성환
    • 한국소음진동공학회:학술대회논문집
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    • 한국소음진동공학회 2012년도 추계학술대회 논문집
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    • pp.300-305
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    • 2012
  • The vibration and stress analysis program of comprehensive vibration assessment program (CVAP) is to verify theoretically the structural integrity of reactor vessel internals (RVI) and to provide the basis for selecting the locations monitored in measurement and inspection programs. This paper covers the verification of the vibration and stress analysis methodology of APR1400 RVI CVAP. The analysis methodology was developed to use 3-dimensional hydraulic and structural models with ANSYS and CFX. To validate the methodology, the hydraulic loads and structural reponses of OPR1000 were predicted and compared with the calculated and measured data in the OPR1000 RVI CVAP. Since the results predicted with this methodology were close to the measured values considerably, it was confirmed that the analysis methodology was developed properly.

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옹벽구조시스템의 신뢰성 및 안전도 해석 (Reliability and Safety Analysis of Structure System of Retaining Walls)

  • 정철원;윤병조
    • 한국구조물진단유지관리공학회 논문집
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    • 제2권3호
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    • pp.223-234
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    • 1998
  • In this study, an attempt is made to apply the concept of fuzzy-bayesian theory to the integrity assessment of structure system, and uncertainty states are represented in terms of fuzzy sets which define several linguistic variables such as "very good", "good", "average", "poor", "very poor", etc. Especially, the concept of fuzzy conditional probability aids to derive a new reliability analysis which includes the subjective assessment of engineers without introducing any additional correction factors. The fuzzy concept are also used as reliability indexes for the condition assessment based on the proposed models, the proposed fuzzy theory-based approach with the results of PEM and AFOSM are applied to retaining wall.

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