• Nuclear Engineering and Technology
• /
• v.38 no.3
• /
• pp.293-300
• /
• 2006

A liquid metal reactor (LMR) operated at high temperatures is subjected to both cyclic mechanical loading and thermal loading; thus, creep-fatigue is a major concern to be addressed with regard to maintaining structural integrity. The Korea Advanced Liquid Metal Reactor (KALIMER), which has a normal operating temperature of $545^{\circ}C$ and a total service life time of 60 years, is composed of various cylindrical structures, such as the reactor vessel and the reactor baffle. This study focuses on the creepfatigue crack initiation for a cylindrical Y-junction structure made of 316 stainless steel (SS), which is subjected to cyclic axial tensile loading and thermal loading at a high-temperature hold time of $545^{\circ}C$. The evaluation of the considered creep-fatigue crack initiation was carried out utilizing the ${\sigma}_d$ approach of the RCC-MR A16 guide, which is the high-temperature defect assessment procedure. This procedure is based on the total accumulated strain during the service time. To confirm the evaluated result, a high-temperature creep-fatigue structural test was performed. The test model had a circumferential through wall defect at the center of the model. The defect front of the test model was investigated after the $100^{th}$ cycle of the testing by utilizing a metallurgical inspection technique with an optical microscope, after which the test result was compared with the evaluation result. This study shows how creep-fatigue crack initiation for a high-temperature structure can be predicted with conservatism per the RCC-MR A16 guide.

• Nuclear Engineering and Technology
• /
• v.43 no.3
• /
• pp.301-308
• /
• 2011

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.05a
• /
• pp.316.2-316.2
• /
• 2004

• Journal of the Korea Concrete Institute
• /
• v.14 no.3
• /
• pp.382-391
• /
• 2002

This study is devoted to the problems of thermal and shrinkage stresses in order to avoid cracking at early ages. The early-age damage induced by volume change has great influence on the long-term structural performance of the concrete structures such as its durability and serviceability To solve this complex problem, the computer programs for analysis of thermal and shrinkage stresses were developed. In these procedures, numerous material models are needed and the realistic numerical models have been developed and validated by comparison with relevant experimental results in order to solve practical problems. A framework has been established for formulation of material models and analysis with 3-D finite element method. After the analysis of the temperature, moisture and degree of hydration field in hardening concrete structure, the stress development is determined by incremental structural formulation derived from the principle of virtual work. In this study, the stress development is related to thermal and shrinkage deformation, and resulting stress relaxation due to the effect of early-age creep. From the experimental and numerical results it is found that the early-age creep p)ays important role in evaluating the accurate stress state. The developed analysis program can be efficiently utilized as a useful tool to evaluate the thermal and shrinkage stresses and to find measures for avoiding detrimental cracking of concrete structures at early ages.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2008.11a
• /
• pp.605-606
• /
• 2008

• Journal of the Korean Geosynthetics Society
• /
• v.3 no.2
• /
• pp.31-38
• /
• 2004

Long-term stability of two type geogrids were evaluated. Membrane drawn type geogrid showed the exponential type tensile property and textile type geogrid showed the rapid increase of tensile property closer toward the break point. Accelerated creep test was done for textile type geogrid but not done for membrane drawn type geogrid because of its thermal property. Creep strain for membrane drawn type geogrid was larger than the ultimate tensile strain by tensile test and reduction factor by creep deformation of textile type geogrid was smaller than that of membrane type geogrid.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.12 no.1
• /
• pp.70-77
• /
• 2016

In this paper, the structural design and integrity evaluations for a reactor vessel of PGSFR sodium-cooled fast reactor(150MWe) are carried out in compliance with ASME BPV III, Division 5 Subsection HB. The reactor vessel is designed with a direct contact of primary sodium coolant to its inner surface and has a double vessel concept enclosing by containment vessel. To assure the structural integrity for 60 years design lifetime and elevated operating temperature of $545^{\circ}C$, which can invoke creep and creep-fatigue damage, the structural integrity evaluations are carried out in compliance with the ASME code rules. The design loads considered in this evaluations are primary loads and operation thermal cycling loads of normal heat-up and cool-down. From the evaluations, the PGSFR reactor vessel satisfies the ASME code limits but it was found that there is a little design margin of creep damage for inner surface at the region of cold pool free surface.

• Structural Engineering and Mechanics
• /
• v.16 no.6
• /
• pp.695-712
• /
• 2003

The non-linear structural analysis of reinforced concrete beams in fire consists of three separate steps: (i) The estimation of the rise of surrounding air temperature due to fire; (ii) the determination of the distribution of the temperature within the beam during fire; (iii) the evaluation of the mechanical response due to simultaneous time-dependent thermal and mechanical loads. Steps (ii) and (iii) are dealt with in the present paper. We present a two-step computational procedure where a 2D transient thermal analysis over the cross-sections of beams are made first, followed by mechanical analysis of the structure. Fundamental to the accuracy of the mechanical analysis is a new planar beam finite element. The effects of plasticity in concrete, and plasticity and viscous creep in steel are taken into consideration. The properties of concrete and steel along with the values of their thermal and mechanical parameters are taken according to the European standard ENV 1992-1-2 (1995). The comparison of our numerical and full-scale experimental results shows that the proposed mechanical and 2D thermal computational procedure is capable to describe the actual response of reinforced concrete beam structures to fire.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 1997.04a
• /
• pp.905-909
• /
• 1997

A geometric error of machine parts is one of the most important factors that affect the accuracy of positioning, generating and measuring for precision machinery. It is known that the thermal deformation of a workpiece during surface grinding is the most important in the geometric error of ground surface. This paper experimentally describes the grinding characteristics of creep-feed grinding. The wheels have 6 slotted pieces in order to compare the grinding temperature with the geometric.

• Journal of the Korean Society of Fisheries and Ocean Technology
• /
• v.28 no.1
• /
• pp.71-78
• /
• 1992

Zr-4 used for a cladding and an end plug of reactor component has creep deformation under operation at high temperature. Creep is regarded as the time dependent deformation of a material under constant applied stress. Although the major source of the deformation of zirconium component in water-cooled reactors is irradiation creep, the thermal creep may give a rise to significant deformation in reactor component especially at relatively high temperatures and at various constant stresses, and therefore it must be predicted accurately. Stress relaxation is the time dependent change of stress at constant strain and it is a process related intimately to creep. In this paper, the creep behavior and stress relaxation of Zr-4 is examined at the temperature of 50$0^{\circ}C$ that is 40% of the absolute melting temperature of Zr-4 under the stress below yield stress and under the various constant strains. The results obtained are summarized as follows: 1) With an increase of stress, the steady state creep rate increases and the creep rupture time decreases. 2) The steady state creep rate $\varepsilon$(%/s) for the stress $\sigma$sub(c) (kgf/mm super(2)) of Zr-4 increases outstandingly. All the empirical equations computed for Zr-4 increases outstandingly. All the empirical equations computed for Zr-4 are in accord with Norton's model equation($\varepsilon$=K$\sigma$ sub(c) super (n)). The constants of materials computed are as follows: K=3.9881$\times$10 super(-5), n=1.9608 3) The rupture time T sub(r) (hr) decreases linearly with the increase of stress on the log-log scaled graph. The empirical equations computed for Zr-4 are in accord with Bailey's model equation (T sub(r)=K sub(1)$\sigma$sub(c) super(m)). The constants of materials computed are as follows: K sub(1)=1.2875$\times$10 super(16), m=-3.467 4) It seems clear that the strain could be quantitatively dependent on the high temperature creep properties such as creep stress, rupture time, steady state creep rate and total creep rate. It is found that these relationships are linear on the log-log graph. 5) In stress relaxation test, as the critical constant strain that can be allowed to the specimen is larger, stress relaxation becomes more rapid, and as the constant strain is smaller, the stress relaxation becomes slower.