• Proceedings of the KSME Conference
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• 2000.11b
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• pp.67-76
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• 2000

Cr-Mo low alloy steels have been used for a long time for pressure vessel due to its excellent corrosion resistance, high temperature strength and toughness. The paper reviewed the latest trends on material development and some problems on Cr-Mo low alloy steel for pressure vessel, such as elevated temperature strength, hardenability, synergetic effect between temper and hydrogen embrittlement, hydrogen attack and hydrogen induced disbonding of overlay weld-cladding.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.67-76
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• 2000

Cr-Mo low alloy steels have been used for a long time for pressure vessel due to its excellent corrosion resistance, high temperature strength and toughness. The paper reviewed the latest trends on material development and some problems on Cr-Mo low alloy steel for pressure vessel, such as elevated temperature strength, hardenability, synergetic effect between temper and hydrogen embrittlement, hydrogen attack and hydrogen induced disbonding of overlay weld-cladding.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.17 no.1
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• pp.153-160
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• 2018

A pressure vessel consists of an inner tank and the outer tank; the material of the inner tank is austenite stainless steel, and the outer tank is general carbon steel. As the storage amount increase, the size of the inner tank for LNG also increases, which eventually increases the weight of the LNG storage tank. The Cold Stretch method can transport and store the LNG in a larger amount than the conventional pressure container, and the weight of the pressure vessel can also be reduced at 50 70% due to the reduction of the thickness, which is excellent from an economic and energy consumption perspective. Although the Cold Stretch method has these advantages, the domestic situation has not developed any related legislation. In this study, the actual production of pressure vessels using the Cold Stretch method will be processed and the volume expansion after the Cold Stretch will be checked and compared with the mechanical properties.

• Proceedings of the KWS Conference
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• 2001.05a
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• pp.228-231
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• 2001

Elastic-plastic fracture toughness $J_{1c}$ can be used as an effective design criterion in elastic plastic fracture mechanics. Most of these systems are operated at high temperature and $J_{1c}$ values are affected by temperature. therefore, the $J_{1c}$ valuse at high temperature must be determined for use of integrity evaluation and designing of such systems. Elastic-plastic fracture toughness $J_{1c}$ tests were performed on SA516 carbon steel plate and test results were analyzed according to ASTM E 813-8, ASTM 1813-89. Safety and integrity are required for reactor pressure vessels vecause pthey are operated in high temperature. there are single specimen method, which used as evaluation of safety and integrity for reactor pressure vessels. In this study, elastic-plastic fracture toughness$(J_{1c})$ and $J-\Delta{a}$ of SA 516/70 steel used as reactor pressure vessel steel are measured and evaluated at room Temperature, $150^{\circ}C$, $250^{\circ}C$ and $370^{\circ}C$ according to unloading compliance method.

• Journal of Ocean Engineering and Technology
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• v.15 no.2
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• pp.99-104
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• 2001

Elastic-plastic fracture toughness $J_{lc}$ can be used as an effective design criterion in elastic plastic fracture mechanics. Most of these systems are$J_{lc}$ $J_{lc}$ value at high temperature must be determined for use of integrity evaluation and designing of such systems. Elastic-plastic fracture toughness $J_{lc}$ tests were performed on SA516/70 carbon steel plate and test results were analyzed according to ASTM E 813-87, ASTM E 813-89 and ASTM E 1152-87.safety and integrity are required for reactor pressure vessels because, they are operated in high temperature. There are single specimen method, which used as evaluation of safety and integrity for reactor pressure vessels. In this study, elastic-plastic fracture toughness($J_{lc}$) and J-$\Delta$a of SA 516/70 steel used as reactor pressure vessel steel are measured and evaluated at room temperature, 150$^{\circ}C $, 250$^{\circ}C $ and 370$^{\circ}C $ according to unloading compliance method.

• Journal of Ocean Engineering and Technology
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• v.14 no.1
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• pp.88-94
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• 2000

Recently various pressure vessels like an atomic reactor and plant facilities become more larger and are needed to bear in both very high temperature and pressure condition. And in making such a high pressure vessels the amount of annual usage of 2 $\frac {1}{4}$ Cr-1Mo steels which are predominant to resist high temperature high pressure and corrosive circumstances are increasing. But despite of this advantage of 2 $\frac {1}{4}$Cr-1Mo steel. when PWHT(post welding heat treatment) is carried out lots of reheating cracks are occur. In this reason it is strongly needed to study and examine the mechanical behavior of welded joints through welding to PWHT process. So in this study welded nozzle of pressure vessel where reheat cracks are frequently occur are selected for analysis the crack-occurrence mechanism.

• Journal of Mechanical Science and Technology
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• v.14 no.9
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• pp.935-946
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• 2000

Elastic-plastic stress analysis has been performed to evaluate the fatigue life of an autofrettaged pressure vessel containing cross-bores subjected to pulsating internal pressure of 200 MPa. Finite element analyses were used to calculate the residual and operating stress distributions of the pressure vessel due to the autofrettage process and pulsating internal pressure, respectively. Theoretical stress concentration factors of 3.06, 2.58, and 2.64 were obtained at the cross-bore of the pressure vessel due to internal pressure, 50%, and 100% autofrettage loadings, respectively. Local stresses and local strains determined from the elastic-plastic finite element analysis were employed to calculate the failure location and fatigue life of the pressure vessel with radial cross-bores, incorporating the low-cycle fatigue properties of the pressure vessel steel and fatigue damage parameters. Increase in the amount of overstrain by autofrettage process moved the crack initiation location from the inner radius toward a mid-wall, and extended the crack initiation life. Predicted fatigue life of the fully autofrettaged pressure vessel with cross-bores increased about 50%, compared to the unautofrettaged pressure vessel. At the autofrettage level higher than 50%, the failure location and fatigue life of the pressure vessel were not significantly influenced by the autofrettage level.

• Nuclear Engineering and Technology
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• v.36 no.3
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• pp.229-236
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• 2004

Radiation hardening is a multiscale phenomenon involving various processes over a wide range of time and length. We present a multiscale model for estimating the amount of radiation hardening in pressure vessel steel in the environment of a light water reactor. The model comprises two main parts: molecular dynamics (MD) simulation and a point defect cluster (PDC) model. The MD simulation was used to investigate the primary damage caused by displacement cascades. The PDC model mathematically formulates interactions between point defects and their clusters, which explains the evolution of microstructures. We then used a dislocation barrier model to calculate the hardening due to the PDCs. The key input for this multiscale model is a neutron spectrum at the inner surface of reactor pressure vessel steel of the Younggwang Nuclear Power Plant No.5. A combined calculation from the MD simulation and the PDC model provides a convenient tool for estimating the amount of radiation hardening.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.132-144
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• 2001

Fracture mechanics is one of the major areas of the pressurized thermal shock (PTS) evaluation. To evaluate the reactor pressure vessel integrity associated with PTS, PFM methodology demands precise calculation of temperature, stress, and stress intensity factor for the variety of PTS transients. However, the existence of stainless steel cladding, with different thermal, physical, and mechanical property, at the inner surface of reactor pressure vessel complicates the fracture mechanics analysis. In this paper, treatment schemes to evaluate stress and resulting stress intensity factor for RPV with stainless steel clad are introduced. For a reference transient, the effects of clad thermal conductivity and thermal expansion coefficients on deterministic fracture mechanics analysis are examined.

• Proceedings of the KSME Conference
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• 2001.11a
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• pp.87-92
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• 2001

In this paper, high strength pressure vessel steels having the same chemical compositions were manufactured by the two different steel-making processes, such as vacuum degassing(VD) and electro-slag remelting(ESR) methods. After the steel-making process, they were normalized at $955^{\circ}C$, quenched at $843^{\circ}C$, and finally tempered at $550^{\circ}C$ or $450^{\circ}C$, resulting in tempered martensitic microstructures with different yielding strengths depending on the tempering conditions. Low-cycle fatigue(LCF) tests, fatigue crack growth rate(FCGR) tests, and fracture toughness tests were performed to investigate the fatigue and fracture behaviors of the pressure vessel steels. In contrast to very similar monotonic, LCF, and FCGR behaviors between VD and ESR steels, a quite difference was noticed in the fracture toughness. Fracture toughness of ESR steel was higher than that of VD steel, being attributed to the removal of impurities in steel-making process.