• Title/Summary/Keyword: Vessel upper head

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ROSA/LSTF test and RELAP5 code analyses on PWR 1% vessel upper head small-break LOCA with accident management measure based on core exit temperature

  • Takeda, Takeshi
    • Nuclear Engineering and Technology
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    • v.50 no.8
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    • pp.1412-1420
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    • 2018
  • An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

Evaluation for Weld Residual Stress and Operating Stress around Weld Region of the CRDM Nozzle in Reactor Vessel Upper Head (원자로 압력용기 상부헤드 CRDM 노즐 용접부의 용접잔류응력 및 운전응력 평가)

  • Lee, Kyoung-Soo;Lee, Sung-Ho;Bae, Hong-Yeol
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.36 no.10
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    • pp.1235-1239
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    • 2012
  • Primary water stress corrosion cracking (PWSCC) has been observed around the weld region of control rod drive mechanism (CRDM) nozzles in nuclear power plants overseas. The weld has a J-shaped groove and it connects the CRDM nozzle with the reactor vessel upper head (RVUH). It is a dissimilar metal weld (DMW), because the CRDM is made of alloy 600 and the RVUH is made of carbon steel. In this study, finite element analysis (FEA) was performed to estimate the stress condition around the weld region. Generally, it is known that a high tensile region is more susceptible to PWSCC. FEA was performed as for the condition of welding, hydrostatic test and normal operation successively to observe how the residual stress changes due to plant condition. The FEA results show that a high tensile stress region is formed around the weld starting point on the inner surface and around the weld stop point on the outer surface.

DYNAMIC CHARACTERISTICS OF A PARTIALLY FLUIDFILLED CYLINDRICAL SHELL

  • Jhung, Myung-Jo;Yu, Seon-Oh;Lim, Yeong-Taek
    • Nuclear Engineering and Technology
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    • v.43 no.2
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    • pp.167-174
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    • 2011
  • A pressurizer in a small integral type pressurized water reactor is located inside the upper region of the reactor vessel, and uses a space between the upper head of the reactor vessel and the upper region of the upper guide structure which is partially filled with fluid depending on the operating power. This new design requires a comprehensive investigation of vibration characteristics. This study investigates the modal characteristics of a pressurizer which uses a simplified cylindrical shell model, focusing on how having fluid in the shell affects vibration and response characteristics. In addition, an analysis of sloshing is performed and the response characteristics are addressed.

Effects of Repair Weld of Reactor Pressure Vessel Upper Head Control Rod Drive Mechanism Penetration Nozzle on J-Groove Weldment Using Finite Element Analysis (유한요소법을 이용한 원자로 상부헤드 CRDM 관통노즐 J-Groove 보수용접 영향 분석)

  • Kim, Ju Hee;Yoo, Sam Hyeon;Kim, Yun Jae
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.38 no.6
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    • pp.637-647
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    • 2014
  • In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite-element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

Root Cause Analysis and Structural Integrity Evaluation for a Crack in a Reactor Vessel Upper Head Penetration Nozzle (원자로 상부헤드 관통노즐 균열에 대한 원인분석 및 건전성 평가)

  • Lee, Kyoung-Soo;Lee, Sung-Ho;Lee, Jeong-Seog;Lee, Jae-Gon;Lee, Seung-Gun
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.9 no.1
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    • pp.56-61
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    • 2013
  • This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

Assessment of CATHARE code against DEC-A upper head SBLOCA experiments

  • Anis Bousbia Salah
    • Nuclear Engineering and Technology
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    • v.56 no.3
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    • pp.866-872
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    • 2024
  • Design Extension Conditions (DEC)-A assessments of the operating nuclear power plants are generally considered for the purpose of getting additional safety demonstrations of their capability to undergo conditions that are generally more severe than DBAs by features implemented in the design and accident management measures. The pursued methodology is generally based upon Best Estimate approaches aiming at verifying that the safety limits in terms of integrity of the barriers against eventual large or early releases of radioactive material are fulfilled. These aspects are nowadays being experimentally and analytically addressed within the OECD/NEA experimental projects like the ATLAS and PKL series where a set of DEC-A experiments are considered. In this paper, experiments related to SBLOCA at the vessel upper head of the pressurized vessel of ATLAS and PKL are analytically assessed using the CATHARE code. These experiments includes issues related to common cause failure of the safety injection system and operator actions for preventing core excessive overheating. It is shown that, on the one hand, the safety features embedded in the design together with the operator actions are capable to prevent the progression towards a severe accident state and on the other hand, the code prediction capabilities for such scenario are generally good but still to be enhanced.

Effect of Normal Operating Condition Analysis Method for Weld Residual Stress of CRDM Nozzle in Reactor Pressure Vessel (원전 정상가동조건 적용 방식이 원자로 압력용기 상부헤드 관통 노즐의 용접 잔류응력에 미치는 영향)

  • Nam, Hyun Suk;Bae, Hong Yeol;Oh, Chang Young;Kim, Ji Soo;Kim, Yun Jae
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.37 no.9
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    • pp.1159-1168
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    • 2013
  • In pressurized water nuclear reactors (PWRs), the reactor pressure vessel (RPV) upper head contains penetration nozzles that use a control rod drive mechanism (CRDM). The penetration nozzle uses J-groove weld geometry. Recently, the occurrence of cracking in alloy 600 CRDM penetration nozzle has increased. This is attributable to primary water stress corrosion cracking (PWSCC). PWSCC is known to be susceptible to the welding residual stress and operational stress. Generally, the tensile residual stress is the main factor contributing to crack growth. Therefore, this study investigates the effect on weld residual stress through different analysis methods for normal operating conditions using finite element analysis. In addition, this study also considers the effect of repeated normal operating condition cycles on the weld residual stress. Based on the analysis result, this paper presents a normal operating condition analysis method.

Identification of nonregular indication according to change of grain size/surface geometry in nuclear power plant (NPP) reactor vessel (RV)-upper head alloy 690 penetration

  • Kim, Kyungcho;Kim, Changkuen;Kim, Hunhee;Kim, Hak-Joon;Kim, Jin-Gyum;Jhung, Myungjo
    • Nuclear Engineering and Technology
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    • v.49 no.7
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    • pp.1524-1536
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    • 2017
  • During the fabrication process of reactor vessel head penetration (RVHP), the grain size of the tube material can be changed by hot or cold work and the inner side of the tube can also be shrunk due to welding outside of the tube. Several nonregular time-of-flight diffraction (TOFD) signals were found because of deformed grains. In this paper, an investigation of nonregular TOFD indications acquired from RVHP tubes using experiments and computer simulation was performed in order to identify and distinguish TOFD signals by coarse grains from those by Primary Water Stress Corrosion Crack (PWSCC). For proper understanding of the nonregular TOFD indications, microstructural analysis of the RVHP tubes and prediction of signals scattered from the grains using Finite Element Method (FEM) simulation were performed. Prediction of ultrasonic signals from the various sizes of side drilled holes to find equivalent flaws, determination of the size of the nonregular TOFD indications from the coarse grains, and experimental investigation of TOFD signals from coarse grain and shrinkage geometry to identify PWSCC signals were performed. From the computer simulation and experimental investigation results, it was possible to obtain the nonregular TOFD indications from the coarse grains in the alloy 690 penetration tube of RVHP; these nonregular indications may be classified as PWSCC. By comparing the computer simulation and experimental results, we were able to confirm a clear difference between the coarse grain signal and the PWSCC signal.

Development of Automated Nondestructive Inspection System for BMI Nozzles in Nuclear Vessel (원자로 BMI 노즐 검사를 위한 자동화 비파괴검사 시스템 개발)

  • Park, Joon Soo;Lee, Won Kun;Han, Won Jin;Lee, Sun Ho;Seong, Un Hak
    • Journal of the Korean Society for Nondestructive Testing
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    • v.33 no.1
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    • pp.26-33
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    • 2013
  • BMI nozzles in bottom head of the nuclear vessel are one of major components in nuclear power plants. The BMI nozzles have high possibility to generate PWSCC(primary water stress corrosion crack) according to recent foreign case although operation temperature is lower then the upper head of the nuclear vessel. Thus, nondestructive inspection of the BMI nozzles is required. But, inspection of BMI nozzles is not easy since the BMI nozzles placed in high radiated area and inside the nozzles filled with boric acid. Thus, in this study, a TOFD transducer for inspection of BMI and automated scanner system with water were developed. Also, validation of performance of the developed transducer and system are performed using specimens with artificial defects.

Saturated Boiling Heat Transfer of Freon-113 in Hemispherical Narrow Space and Implications for Degraded Core Coolability in Reactor Vessel Lower Plenum

  • Bang, Kwang-Hyun
    • Proceedings of the Korean Nuclear Society Conference
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    • 1995.05a
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    • pp.574-579
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    • 1995
  • Saturated boiling heat transfer experiment in a hemispherical narrow space is conducted using Freon-113 to investigate an additional heat removal capability through a hypothetical gap between lower head and degraded core. The narrow space of 1mm consists of a 124mm diameter heated stainless steel hemisphere and a glass outer vessel. Within the hemispherical narrow space large coalesced bubbles are produced and these bubbles rise in random direction, causing liquid flow in from the opposite side to fill the region. Such flow in random direction makes the flow field in the narrow space very chaotic and thus enhance heat transfer. The heat transfer coefficient is higher at lower angle and at higher heat flux. The present study shows that the liquid from upper region can effectively penetrate into the gap and augment the heat removal capability through tile gap.

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