• Title/Summary/Keyword: nuclear reactor vessel steel

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Investigation of a Hydrogen Mitigation System During Large Break Loss-Of-Coolant Accident for a Two-Loop Pressurized Water Reactor

  • Dehjourian, Mehdi;Sayareh, Reza;Rahgoshay, Mohammad;Jahanfarnia, Gholamreza;Shirani, Amir Saied
    • Nuclear Engineering and Technology
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    • v.48 no.5
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    • pp.1174-1183
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    • 2016
  • Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

Quantitative Estimation of Radiation Damage in Reactor Pressure Vessel Steels by Using Multiscale Modeling (멀티스케일 모델링을 이용한 압력용기강의 조사손상 정량예측)

  • Lee, Gyeong-Geun;Kwon, Junhyun
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.10 no.1
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    • pp.113-121
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    • 2014
  • In this work, an integrated model including molecular dynamics and chemical rate theory was implemented to calculate the growth of point defect clusters(PDC) and copper-rich precipitates(CRP) which could change the mechanical properties of reactor pressure vessel(RPV) steels in a nuclear power plant. A number of time-dependent differential equations were established and numerically integrated to estimate the evolution of irradiation defects. The calculation showed that the concentration of the vacancies was higher than that of the self-interstitial atoms. The higher concentration of vacancies induced a formation of the CRPs in the later stage. The size of the CRPs was used to estimate the mechanical property changes in RPV steels, as is the same case with the PDCs. The calculation results were compared with the measured values of yield strength change and Charpy V-notch transition temperature shift, which were obtained from the surveillance test data of Korean light water reactors(LWRs). The estimated values were in fair agreement with the experimental results in spite of the uncertainty of the modeling parameters.

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.

Effects of Microstructural States on Magnetic Barkhausen Noise Behavior in the Weld Heat-Affected Zone of Reactor Pressure Vessel Steel (원자로압력용기강 용접열영향부의 미세조직 변화가 Magnetic Barkhausen Noise 거동에 미치는 영향)

  • Kim, Joo-Hag;Yoon, Eui-Pak;Moon, Jong-Gul;Park, Duck-Gun;Hong, Jun-Hwa
    • Journal of the Korean Society for Nondestructive Testing
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    • v.18 no.4
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    • pp.292-303
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    • 1998
  • Recent study has demonstrated that some magnetic properties are sensitive to the microstructural state of material. The ASTM A 508 Gr. 3 reactor pressure vessel steel has various microstructural changes including martensitic and bainitic phases, and various sizes of grain and precipitates in the weld heat-affected zone (HAZ). To correlate the microstructural state with Barkhausen noise (BN), specimens were prepared through simulating various weld thermal cycles using a thermal simulator. The conventional magnetic properties, i.e. coercive force, remanence and maximum induction, did not change significantly, whereas the BN amplitude and energy during a magnetization cycle changed markedly with microstructural state. The BN increased with increasing grain and carbide sizes, and the tempered bainite structure showed higher BN parameter than tempered martensite.

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Effect of Neutron Energy Spectra on the Formation of the Displacement Cascade in ${\alpha}-Iron$

  • Kwon Junhyun;Seo Chul Gyo;Kwon Sang Chul;Hong Jun-Hwa
    • Nuclear Engineering and Technology
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    • v.35 no.5
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    • pp.497-505
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    • 2003
  • This paper describes a computational approach to the quantification of primary damage under irradiation and demonstrates the effect of neutron energy spectra on the formation of the displacement cascade. The development of displacement cascades in ${\alpha}-Iron$ has been simulated using the MOLDY code - a molecular dynamics code for simulating radiation damage. The primary knock-on atom energy, key input to the MOLDY code, was determined from the SPECTER code calculation on two neutron spectra. The two neutron spectra include; (i) neutron spectrum in the instrumented irradiation capsule of the high-flux advanced neutron application reactor (HANARO), and (ii) neutron spectrum at the inner surface of the reactor pressure vessel steel for the Younggwang nuclear power plant No.5 (YG 5). Minor differences in the normalized neutron spectra between the two spectra produce similar values of PKA energy, which are 4.7 keV for HANARO and 5.3 keV for YG 5. This similarity implies that primary damage to the components of the commercial nuclear reactors should be well simulated by irradiation in the HANARO. Moreover, the application of the MD calculations corroborates this statement by comparing cascades simulation results.

Effects of Ni and Cr Contents on the Fracture Toughness of Ni-Mo-Cr Low Alloy Steels in the Transition Temperature Region (Ni-Mo-Cr계 저합금강의 천이온도영역에서의 파괴인성에 미치는 Ni 및 Cr 함량의 영향)

  • Lee, Ki-Hyoung;Park, Sang-Gyu;Kim, Min-Chul;Lee, Bong-Sang;Wee, Dang-Moon
    • Korean Journal of Metals and Materials
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    • v.47 no.9
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    • pp.533-541
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    • 2009
  • Materials used for a reactor pressure vessel(RPV) are required high strength and toughness, which determine the safety margin and life of a reactor. Ni-Mo-Cr low alloy steel shows better mechanical properties than existing RPV steels due to higher Ni and Cr contents compared to the existing RPV steels. The present study focuses on effects of Ni, Cr contents on the cleavage fracture toughness of Ni-Mo-Cr low alloy steels in the transition temperature region. The fracture toughness was characterized by a 3-point bend test of precracked Charpy V-notch(PCVN) specimens based on ASTM E1921-08. The test results indicated that the fracture toughness was considerably improved with an increase of Ni and Cr contents. Especially, control of Cr content was more effective in improving fracture toughness than manipulating Ni content, though Charpy impact toughness was changed more extensively by adjusting Ni content. These differences between changes in the fracture toughness and that in the impact toughness were derived from microstructural features, such as martensite lath size and carbide precipitation behavior.

A NEW BOOK: 'LIGHT-WATER REACTOR MATERIALS'

  • OLANDER DONALD R.;MOTTA ARTHUR T.
    • Nuclear Engineering and Technology
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    • v.37 no.4
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    • pp.309-316
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    • 2005
  • The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the $UO_2$ fuel, Zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed.

Welding process for manufacturing of Nuclear power main components (원자력 발전 주기기 제작에 적용되는 용접공정)

  • Jung, In-Chul;Kim, Yong-Jae;Shim, Deog-Nam
    • Proceedings of the KWS Conference
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    • 2010.05a
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    • pp.43-46
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    • 2010
  • As the nuclear power plant has been constructed continuously for several decades in Korea, the welding technology for components manufacturing and installation has been improved largely. Standardization for weld test and qualification was also established systematically according to the concerned code. The welding for the main components requires the high reliability to keep the constant quality level, which means the repeatability of weld quality. Therefore the weld process qualified by thorough test and evaluation is able to be applied for manufacturing. Narrow gap SAW and GTAW process are usually applied for girth seam welding of pressure vessel like Reactor vessel, steam generator, and etc. For the surface cladding with stainless steel and Inconel material, strip welding process is mainly used. Inside cladding of nozzles is additionally applied with Hot wire GTAW and semi-auto welding process. Especially the weld joint having elliptical weld line on curved surface needs a specialized weld system which is automatically rotating with adjusting position of the head torch. The small sized pipe, tube, and internal parts of reactor vessel requests precise weld processes like an automatic GTAW and electron beam welding. Welding of dissimilar materials including Inconel690 material has high possibility of weld defects like a lack of fusion, various types of crack. To avoid these kinds of problem, optimum weld parameters and sequence should be set up through the many tests. As the life extension of nuclear power plant is general trend, weld technologies having higher reliability is required gradually. More development of specialized welding systems, weld part analysis and evaluation, and life prediction for main components should be taken into a consideration extensively.

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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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A Strategy for Kori Unit 1 Pressure Vessel Fluence Reduction through a Modification of Outer Assembly Configuration Using Monte Carlo Analysis

  • Kim, Jae-Cheon;Kim, Jong-Kyung;Kim, Jong-Oh
    • Proceedings of the Korean Nuclear Society Conference
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    • 1997.05b
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    • pp.515-519
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    • 1997
  • The purpose of this study is to reduce the fast neutron fluence at the reactor pressure vessel(RPV) and to provide a basis for plant-life extension. In this study, different neutron absorbers were employed in the core outer assemblies of Kori Unit 1 Cycle 14. The modified assemblies were used to calculate fast neutron fluence at the RPV and to evaluate reduction of outer assembly power and total power in core. By comparison with the case of no suppression fixture, the fast neutron fluence of a case with two rows stainless steel around the assembly with natural uranium pins is decreased by 85.8%. It is noted that the modification of outer assembly is more efficient than the previous low leakage loading pattern (LLLP) applied to Kori Unit 1. Also, compared fast neutron fluence in Cycle 1 with Cycle 14, fast neutron fluence at the RPV between Cycle 1 and Cycle 14 is not significantly different. It is found that LLLP applied to the Kori Unit 1 has not contributed to fast neutron fluence reduction at the RPV.

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