• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05b
• /
• pp.175-180
• /
• 1998

The U-Zr metallic alloy with low swelling HT9 cladding is the candidate for the KALIMER fuel rod. The fuel rod should be able to maintain the structural integrity during its lifetime in the reactor. In a typical metallic fuel rod, load is mainly applied by internal gas pressure, and the deformation is primarily caused by creep of the cladding. The three-dimensional FEM modelling of a fuel rod is important to predict the structural behavior in concept design stage. Using the ANSYS code, the 3-D structure analyses were performed for various configuration, element and loads. It has been shown that the present analysis model properly evaluate the structural integrity of fuel rod. The present analysis results show that the fuel rod is expected to maintain its structural integrity during normal operation.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
• /
• 2008.11a
• /
• pp.457-458
• /
• 2008

We have developed new process to product Al clad steel wire. New machine was modified to be able to apply an four step of "foiling-sizing-cladding-drawing" considering low clad temperature and high clad pressure. The foiling part for continuous foiling of Al sheet was designed and machine. Cladding properties at Al and steel interface were investigated for the processes of new work.

• Transactions of Materials Processing
• /
• v.18 no.1
• /
• pp.80-86
• /
• 2009

This paper presents the plastic inhomogeneous deformation behavior of bimetal composite rods during the axisymmetric and steady-state extrusion process through a conical die. The rigid-plastic FE model considering frictional contact problem was used to analyze the co-extrusion process with material combinations of Cu/Al. Different cases of initial geometry shape for composite material were simulated under different conditions of co-extrusion process, which includes the interference and frictional conditions. From the simulation results, the sleeve cladding rate at the core/sleeve interface was recorded as a distribution of diameter ratio and interference conditions, which will be useful for the investigations of the bonding process during co-extrusion process. In addition, the results of the co-extrusion, connected with the results of the variations of diameter rate and average contact pressure, demonstrate a good agreement and present the possibility of describing the parameters of the plastic zones in non-uniform deformation of these type of composite materials.

• Journal of Welding and Joining
• /
• v.32 no.2
• /
• pp.4-8
• /
• 2014

In this research, GTA overlay welding system is developed for inside of straight pipes in various diameter. It can be applied to oil, ship building and plant industry, especially pipes connected to pressure vessels, for the purpose of cost reduction by cladding inside of pipes with corrosion and heat resistant alloys such as stainless steel or Inconel. Developed system consists of GTA power source, torch, wire feeding system, automatic arc length adjusting device, CCD camera and cooling unit. Two types of pipe inside overlay welding system are developed. One is for maximum 3m pipe length with 3 inch ~ 12 inch pipe outer diameter. Another type can be applied to maximum 12m pipe length with 7 ~ 24 inch OD. Developed system successfully produced inside cladded pipe and the results are shown through cross sectional images of the pipes.

• Nuclear Engineering and Technology
• /
• v.50 no.2
• /
• pp.306-312
• /
• 2018

There is an increasing need in the United States and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations or interim storage sites. Under normal conditions, the Nuclear Regulatory Commission limits cladding temperature to $400^{\circ}C$ for high-burnup (>45 GWd/mtU) fuel, with higher temperatures allowed for low-burnup fuel. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at $400^{\circ}C$. Results were representative of the majority of US light water reactor fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

• Nuclear Engineering and Technology
• /
• v.34 no.4
• /
• pp.286-300
• /
• 2002

An algorithm for the sodium boiling model has been developed for calculation of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. Modeling of sodium boiling in liquid metal reactors using sodium as a coolant is necessary because of phenomenon difference comparing with that observed generally in light water reactor systems. The applied model to the algorithm is the multiple-bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbies that (ill the whole cross section of the coolant channel except for the liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble The present study is focused on not only demonstration of the vapor bubble behavior predicted by the developed model, but also confirmation of a qualitative acceptance for the model. As a result, the model can represent important phenomena in the sodium boiling, but it is found that further effort is also needed for its completition.

• Nuclear Engineering and Technology
• /
• v.49 no.5
• /
• pp.1031-1043
• /
• 2017

A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

• Nuclear Engineering and Technology
• /
• v.50 no.8
• /
• pp.1412-1420
• /
• 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.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.40 no.5
• /
• pp.437-447
• /
• 2016

In a nuclear power plant, the fuel assembly, which is composed of fuel rods, burns, and the high temperature can generate power. The fuel rod consists of pellets and a cladding that covers the pellets. It is important to understand the pellet-cladding mechanical interaction with regard to nuclear safety. This paper proposes simulation of the PCMI. The gap between the pellets and the cladding, and the contact pressure are very important for conducting thermal analysis. Since the gap conductance is not known, it has to be determined by a suitable method. This paper suggests a solution. In this study, finite element (FE) contact analysis is conducted considering thermal expansion of the pellets. As the contact causes plastic deformation, this aspect is considered in the analysis. A 3D FE module is developed to analyze the PCMI using FORTRAN 90. The plastic deformation due to the contact between the pellets and the cladding is the major physical phenomenon. The simple analytical solution of a cylinder is proposed and compared with the fuel rod performance code results.

• Proceedings of the KSME Conference
• /
• 2007.05b
• /
• pp.2488-2493
• /
• 2007

A gap conductance is very important factor which can affect nuclear fuel temperature. Especially, in case of an annular fuel, a gap conductance effect can lead an unexpected heat split phenomena which is caused by a large difference of an inner and outer gap conductance. The gap conductance mechanism is very complicated behavior due to the its strong dependency on microscopic factors such as a contact surface roughness, local contact pressure and local temperature. In this paper, for the decision of test temperature and pressure range, a procedure and calculation results of in-reactor fuel temperature and pressure analysis are summarized which can be applied to test equipment design and determination of test matrix. Based upon analysis results, it is concluded that the minimum and maximum test temperature are $300^{\circ}C$ and $530^{\circ}C$ respectively, and the maximum pellet/cladding interfacial contact pressure should be observed up to 45MPa.