• Title/Summary/Keyword: pressurized vessel

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Stress Intensity factor Calculation for the Axial Semi-Elliptical Surface Flaws on the Thin-Wall Cylinder Using Influence Coefficients (영향계수를 이용한 원통용기 축방향 표면결함의 응력확대계수의 계산)

  • Jang, Chang-Heui;Moon, Ho-Rim;Jeong, Ill-Seok;Kim, Tae-Ryong
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.26 no.11
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    • pp.2390-2398
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    • 2002
  • For integrity analysis of nuclear reactor pressure vessel, including the Pressurized thermal shock analysis, the fast and accurate calculation of the stress intensity factor at the crack tip is needed. For this, a simple approximation scheme is developed and the resulting stress intensity factors for axial semi-elliptical cracks in cylindrical vessel under various loading conditions are compared with those of the finite element method and other approximation methods, such as Raju-Newman's equation and ASME Sec. Xl approach. For these, three-dimensional finite-element analyses are performed to obtain the stress intensity factors for various surface cracks with t/R = 0.1. The approximation methods, incorporated in VINTIN (Vessel INTegrity analysis-INner flaws), utilizes the influence coefficients to calculate the stress intensity factor at the crack tip. This method has been compared with other solution methods including 3-D finite clement analysis for internal pressure, cooldown, and pressurized thermal shock loading conditions. The approximation solutions are within $\pm$2.5% of the those of FEA using symmetric model of one-forth of a vessel under pressure loading, and 1-3% higher under pressurized thermal shock condition. The analysis results confirm that the VINTIN method provides sufficiently accurate stress intensity factor values for axial semi-elliptical flaws on the surface of the reactor pressure vessel.

Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

  • Leskovar, Matjaz;Ursic, Mitja
    • Nuclear Engineering and Technology
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    • v.48 no.1
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    • pp.72-86
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    • 2016
  • A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel-coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.

Analysis of Chemistry Factor and RTPTS Margin for Domestic Reactor Pressure Vessel Materials by using the Surveillance Data (감시시험 결과를 이용한 국내원전 압력용기 재료의 Chemistry Factor 및 RTPTS 평가여유도 분석)

  • Lee, Ho-Jin;Yoon, Ji-Hyun;Choi, Kwon-Jae;Lee, Bong-Sang
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.7 no.3
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    • pp.15-22
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    • 2011
  • The chemistry factor and RTPTS margin for domestic reactor pressure vessel materials were analyzed by using the surveillance data which have been obtained from 8 nuclear power plants in Korea. The surveillance data have been used to assess the integrity of the pressure vessel under the pressurized thermal shock (PTS) event. The chemistry factor, which is determined by the Cu and Ni contents of vessel materials, is considered a proper tool to assess the $RT_{PTS}$. The chemistry factors, which were obtained from the surveillance data of domestic reactor pressure vessels, were investigated and compared with those of Regulatory Guide 1.99 in this study. Regressions for ${\Delta}RT_{NDT}$ were performed to expect the chemistry factor as a function of Cu and Ni, and to estimate $RT_{PTS}$ margin. The margin analysis was performed by comparing the regression graphs and standard deviations with those of Regulatory Guide 1.99. The standard deviations calculated by using the domestic surveillance data for base metal and welds are almost same as the standard deviations which are suggested on Regulatory Guide 1.99, Rev.2.

Vessel failure sensitivities of an advanced reactor for SBLOCA

  • Jhung, Myung Jo;Oh, Chang-Sik;Choi, Youngin;Kang, Sung-Sik
    • Nuclear Engineering and Technology
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    • v.52 no.1
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    • pp.185-191
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    • 2020
  • Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×1019 n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.

Analysis of the flow distribution and mixing characteristics in the reactor pressure vessel

  • Tong, L.L.;Hou, L.Q.;Cao, X.W.
    • Nuclear Engineering and Technology
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    • v.53 no.1
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    • pp.93-102
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    • 2021
  • The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

Constraint-corrected fracture mechanics analysis of nozzle crotch corners in pressurized water reactors

  • Kim, Jong-Sung;Seo, Jun-Min;Kang, Ju-Yeon;Jang, Youn-Young;Lee, Yun-Joo;Kim, Kyu-Wan
    • Nuclear Engineering and Technology
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    • v.54 no.5
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    • pp.1726-1746
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    • 2022
  • This paper presents fracture mechanics analysis results for various cracks located at pressurized water reactor pressure vessel nozzle crotch corners taking into consideration constraint effect. Technical documents such as the ASME B&PV Code, Sec.XI were reviewed and then a fracture mechanics analysis procedure was proposed for structural integrity assessment of various nozzle crotch corner cracks under normal operation conditions considering the constraint effect. Linear elastic fracture mechanics analysis was performed by conducting finite element analysis with the proposed analysis procedure. Based on the evaluation results, elastic-plastic fracture mechanics analysis taking into account the constraint effect was performed only for the axial surface crack of the reactor pressure vessel outlet nozzle with cladding. The fracture mechanics analysis result shows that only the axial surface crack in the reactor pressure vessel outlet nozzle has the stress intensity factor exceeding the low bound of upper-shelf fracture toughness irrespectively of considering the constraint effect. It is confirmed that the J-integral for the axial crack of the outlet nozzle does not exceed the ductile crack initiation toughness. Hence, it can be ensured that the structural integrity of all the cracks is maintained during the normal operation.

PROPOSAL FOR DUAL PRESSURIZED LIGHT WATER REACTOR UNIT PRODUCING 2000 MWE

  • Kang, Kyoung-Min;Noh, Sang-Woo;Suh, Kune-Yull
    • Nuclear Engineering and Technology
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    • v.41 no.8
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    • pp.1005-1014
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    • 2009
  • The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

A Study on the Integrity Evaluation Method of Subclad Crack under Pressurized Thermal Shock (가압열충격 사고시 클래스 하부균열 안전성 평가 방법에 관한 연구)

  • Koo, Bon-Geol;Kim, Jin-Su;Choi, Jae-Boong;Kim, Young-Jin
    • Proceedings of the KSME Conference
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    • 2000.11a
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    • pp.286-291
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    • 2000
  • The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and number of subclad cracks have been found during an in-service-inspection. Therefore assessment for subclad cracks should be made for normal operating conditions and faulted conditions such as PTS. Thus, in order to find the optimum fracture assessment procedures for subclad cracks under a pressurized thermal shock condition, in this paper, three different analyses were performed, ASME Sec. XI code analysis, an LEFM(Liner elastic fracture mechanics) analysis and an EPFM(Elastic plastic fracture mechanics) analysis. The stress intensity factor and the Maximum $RT_{NDT}$ were used for characterizing. Analysis based on ASME Sec. XI code does not completely consider the actual stress distribution of the crack surface, so the resulting Maximum allowable $RT_{NDTS}$ can be non-conservative, especially for deep cracks. LEFM analysis, which does not consider elastic-plastic behavior of the clad material, is much more non-conservative than EPFM analysis. Therefore, It is necessary to perform EPFM analysis for the assessment of subclad cracks under PTS.

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Effect of Flaw Characterization on the Structural Integrity Evaluation Under Pressurized Thermal Shock (가압열충격 사고시 결함 이상화 방법이 구조물 건전성 평가에 미치는 영향)

  • Kim, Jin-Su;Choe, Jae-Bung;Kim, Yeong-Jin;Park, Yun-Won
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.25 no.2
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    • pp.275-282
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    • 2001
  • The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement. Number of subclad cracks may be found during an in-service-inspection due to the presence of cladding. It is specified, in ASME Sec. XI, that a subclad crack is characterized as a surface crack when the thickness of the clad is less than 40% of the crack depth. This condition is provided to keep the crack integrity evaluation conservative. In order to refine the fracture assessment procedures for such subclad cracks under a pressurized thermal shock condition, three dimensional finite element analyses are applied for various subclad cracks existing under cladding. A total of 36 crack geometries are analyzed, and the results are compared with those for surface cracks. The resulting stress intensity factors for subclad cracks are 6 to 44% less than those for surface cracks. It is proven that the flaw characterization condition as specified in ASME Sec. XI can be overly conservative for some subclad cracks.

THE EFFECT OF POSTULATED FLAWS ON THE STRUCTURAL INTEGRITY OF RPV DURING PTS

  • Jhung, Myung-Jo;Choi, Young-Hwan;Chang, Yoon-Suk;Kim, Jong-Wook
    • Nuclear Engineering and Technology
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    • v.39 no.5
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    • pp.647-654
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    • 2007
  • Postulation of flaws, one of the most important areas in RPV integrity assessment, significantly affects the results. In the present work, several parameters, such as orientation, underclad vs. surface cracking, crack depth and shape, etc., are postulated and parametric studies are performed to investigate the influence of the flaw parameters on the structural integrity assessment of the reactor pressure vessel during pressurized thermal shock. The influence of individual parameters describing the crack is evaluated based on sensitivity study results.