• Journal of Welding and Joining
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• v.13 no.2
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• pp.139-149
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• 1995

Many pressure vessels for the power plants are fabricated from low alloy ferritic steels. The inner sides of the pressure vessels are commonly weld_cladded with austenitic stainless steels to minimize problems of corrosive attack. The submerged-arc welding(SAW) process is now used in preference to other processes because of the possibilities open to automation to reduce the overaII welding times. The most reliable way to avoid underclad cracks(UCC) which are often detected at the overlap of the clad beads is to use nonsusceptible steels such as SA508 class 3. At present domestically developed forging steel of SA508 cl.S is now being cladded with single layer by using 90mm wide strip, which transfers higher heat input into the base metal compared to the conventional two layers strip cladding which has been in wide use with 30-60 mm wide strip. But the current indices for the influence of heat input on crack susceptibility are not accurate enough to express the subtle difference in crack susceptibility of the steel. Therefore, the purpose of this present study is: l) To determine UCC susceptibility on domestic forging steel, SA508 cl.S cladded with single layer by using submerged arc 90mm strip and, 2) To optimize heat input range by which the crack susceptibility could be eliminated.

• 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.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2005.05a
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• pp.303-306
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• 2005

Two composites of five plies of STS/Al/Al/Al/STS and STS/Al/STS/Al/STS were produced by roll-cladding at $350^{\circ}C$ from ferritic stainless steel (STS) and aluminum (Al) sheets. In order to analyze the strain states during roll-cladding, the evolution of textures at different through-thickness positions in the roll-clad composites was investigated. Simulations with the finite element method (FEM) disclosed that a strain state which was similar to that of normal rolling with a high friction between roll surface and Al sample led to the formation of texture gradients in the Al sheets in the STS/Al/Al/Al/STS composite. Differences in the material velocity of STS and Al in the rolling direction gave rise to the formation of the shear texture in the Al sheets in the STS/Al/STS/Al/STS composite.

• Journal of Powder Materials
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• v.20 no.4
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• pp.269-274
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• 2013

The present work investigated the dispersion behavior of $Y_2O_3$ particles into AISI 316L SS manufactured using laser cladding technology. The starting particles were produced by high energy ball milling in 10 min for prealloying, which has a trapping effect and homogeneous dispersion of $Y_2O_3$ particles, followed by laser cladding using $CO_2$ laser source. The phase and crystal structures of the cladded alloys were examined by XRD, and the cross section was characterized using SEM. The detailed microstructure was also studied through FE-TEM. The results clearly indicated that as the amount of $Y_2O_3$ increased, micro-sized defects consisted of coarse $Y_2O_3$ were increased. It was also revealed that homogeneously distributed spherical precipitates were amorphous silicon oxides containing yttrium. This study represents much to a new technology for the manufacture and maintenance of ODS alloys.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.7
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• pp.1139-1146
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• 2001

The reactor pressure vessel(RPV) is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and a number of subclad cracks have been found during an in-service-inspection. These subclad cracks should be assured for a safe operation under normal conditions and faulted conditions such as pressurized thermal shock(PTS). Currently available integrity assessment procedure for an RPV, ASME Code Sec. XI, are built on the basis of linear fracture mechanics (LEFM). In PTS condition, however, thermal stress and mechanical stress give rise to high tensile stress at the cladding and elastic-plastic behavior is expected in this area. Therfore, ASME Code Sec. XI is overly conservative in assessing the structural integrity under PTS condition. In this paper, the fracture parameter (stress intensity factor, K, and RT(sub)NDT) from elastic analysis using ASME Sec. XI and finite element method were validated against 3-D elastic-plastic finite element analyses. The difference between elastic and elastic-plastic analysis became significant with increasing crack depth. Therfore, it is recommended to perform elastic-plastic analysis for the accurate assessment of subclad cracks under TPS which causes plastic deformation at the cladding.

• Korean Journal of Metals and Materials
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• v.47 no.9
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• pp.573-579
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• 2009

Titanium/aluminum/stainless steel(Ti/Al/STS) clad materials have received much attention due to their high specific strength and corrosion-resisting properties. However, it is difficult to fabricate these materials, because titanium oxide is easily formed on the titanium surface during heat treatment. The aim of the present study is to derive optimized cladding conditions and thereupon obtain the stable quality of Ti/Al/STS clad materials. Ti sheets were prepared with and without pre-heat treatment and Ti/Al/STS clad materials were then fabricated by cold rolling and a post-heat treatment process. Microstructure of the Ti/Al and STS/Al interfaces was observed using a Scanning Electron Microscope(SEM) and an Energy Dispersed X-ray Analyser(EDX) in order to investigate the effects of Ti pre-heat treatment on the bond properties of Ti/Al/STS clad materials. Diffusion bonding was observed at both the Ti/Al and STS/Al interfaces. The bonding force of the clad material with non-heat treated Ti was higher than that with pre-heat treated Ti before the cladding process. The bonding force decreased rapidly beyond $400^{\circ}C$, because the formed Ti oxide inhibited the joining process between Ti and Al. Bonding forces of STS/Al were lower than those of Ti/Al, because brittle $Fe_3Al$, $Al_3Fe$ intermetallic compounds were formed at the interface of STS/Al during the cladding process. In addition, delamination of the clad material with pre-heat treated Ti was observed at the Ti/Al interface after a cupping test.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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• pp.39-47
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• 2002

Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock(PTS). The PTS event means an event or transient in pressurized water reactors(PWRs) causing severe overcooling(thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored.

• 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.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• 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.