• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.202-202
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• 2002

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.908-913
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• 2000

In the process of integrity evaluation for nuclear power plant components, a series of fracture mechanics evaluation on surface cracks in reactor pressure vessel(RPV) must be conducted. These fracture mechanics evaluations are based on stress intensity factor, K. However, under pressurized thermal shock(PTS) conditions, the combination of thermal and mechanical stress by steep temperature gradient and internal pressure causes considerably high tensile stress at the inside of RPV wall. Besides, the internal pressure during the normal operation produces high tensile stress at the RPV wall. As a result cracks on inner surface of RPVs may experience elastic-plastic behavior which can be explained with J-integral. In such a case, however, J-integral may possibly lose its validity due to constraint effect. In this paper, in order to verify the suitability of J-integral, two dimensional finite element analyses were applied for various surface crack. Total of 18 crack geometries were analyzed, and Q stresses were obtained by comparing resulting HRR stress distribution with corresponding actual stress distributions. In conclusion, HRR stress fields were found to overestimate the actual crack-tin stress field due to constraint effect.

• 한국연소학회:학술대회논문집
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• 2005.10a
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• pp.110-119
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• 2005

The combustion characteristics of binary component single droplets hanging at the tip of a quartz fiber are studied experimentally at different environmental pressures and temperatures under normal gravity. Normal Heptane and Normal Hexadecane are selected as two fuels with high difference in boiling temperatures. A falling electrical furnace in a high pressure vessel has provided high temperature environment. Nitrogen and air have formed the environment to study evaporation and combustion, respectively. The initial diameter of droplet was ranging from 1.1 to 1.3 mm. The evaporation and combustion processes were recorded by a high speed digital camera. Some characteristics of droplet burning under different environment conditions and different droplet composition have been investigated. Microexplosion of droplet take places under atmospheric pressure. Bubble formation and its consequent result, incomplete droplet disintegration which presents in all binary compositions, do not appear at high pressure. The initiation of combustion, always takes place in the bottom of droplet due to buoyancy effect of relatively cold fuel vapor. Also, the burning of binary droplet produces soot when the pressure is high.

• Korean Journal of Metals and Materials
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• v.48 no.2
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• pp.122-132
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• 2010

Higher strength and fracture toughness of reactor pressure vessel steels can be obtained by changing the material specification from that of Mn-Mo-Ni low alloy steel (SA508 Gr.3) to Ni-Mo-Cr low alloy steel (SA508 Gr.4N). However, the operation temperature of the reactor pressure vessel is more than $300^{\circ}C$ and the reactor operates for over 40 years. Therefore, we need to have phase stability in the high temperature range in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel. It is very important to evaluate the temper embrittlement phenomena of SA508 Gr.4N for an RPV application. In this study, we have performed a Charpy impact test and tensile test of SA508 Gr.4N low alloy steel with changing impurity element contents such as Mn and P. And also, the mechanical properties of these low alloy steels after longterm heat treatment ($450^{\circ}C$, 2000hr) are evaluated. Further, evaluation of the temper embrittlement by fracture analysis was carried out. Temper embrittlement occurs in KL4-Ref and KL4-P, which show a decrease of the elongation and a shifting of the transition curve toward high temperature. The reason for the temper embrittlement is the grain boundary segregation of the impurity element P and the alloying element Ni. However, KL4-Ref shows temper embrittlement phenomena despite the same contents of P and Ni compared with SC-KL4. This result may be caused by the Mn contents. In addition, the behavior of embrittlement is not largely affected by the formation of $M_3P$ phosphide or the coarsening of Cr carbides.

• Journal of Ocean Engineering and Technology
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• v.14 no.1
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• pp.44-51
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• 2000

Safety and integrity are required for reactor pressure vessels because they are operated in high temperature. There are single specimen method multiple specimen method and load ratio analysis method which used as evaluation of safety and integrity for reactor pressure vessels. In this study the fracture resistance curve(J-R curve) elastic-plastic fracture toughness($J_{IC}$) and material tearing modulus ($T_{mat}$) of SA 508 class 3 alloy steel used as reactor pressure vessel steel are measured and evaluated at room temperature 20$0^{\circ}C$ and 30$0^{\circ}C$ according to unloading compliance method and load ration analysis method. And then the comparison with experimental $J_{IC}$ and theoretical$J_{IC}$ by local fracture strain is managed.

• Journal of Ocean Engineering and Technology
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• v.15 no.2
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• pp.105-110
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• 2001

The fatigue crack propagation behavior of the SA516/70 steel which is used for pressure vessels was examined experimentally at room temperature, 150$^{\circ}C $, 250$^{\circ}C $ and 370$^{\circ}C $ with stress ratio of R=0.1 and 0.3. The fatigue crack propagation rate da/dN related with the stress intensity factor range $\Delta K$ was influenced by the stress ratio within the stable growth of fatigue crack(Region II) with an increase in $\Delta K$. The resistance to the fatigue crack growth at high temperature is higher in comparison with that at room temperature, and the resistance attributed to the extent of plasticity-induced by compressive residual stress according to the cyclic loads. Fractographic examinations reveal that the differences of the fatigue crack growth characteristics between room and high temperature are mainly explained by the crack closure and oxide-induced by high temperature.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.717-724
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• 2008

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting the reactor pressure vessel and the intermediate heat exchanger for the VHTR. In this study, structural sizing methodology for the primary HGD with a coaxial double-tube of the VHTR that produces heat at temperatures in the order of $950^{\circ}C$ was suggested and a structural pre-sizing of it was carried out as an example.

• Proceedings of the Korea Committee for Ocean Resources and Engineering Conference
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• 2000.10a
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• pp.147-153
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• 2000

The fatigue crack propagation behavior of the SA516/70 steel which is used for pressure vessels was examined experimentally at room temperature, $150^{\circ}C$, $250^{\circ}C$ and $370^{\circ}C$ with stress ratio of R=0.1 and 0.3. The fatigue crack propagation rate da/dN related with the stress intensity factor range $\Omega\textrm{K}$ was influenced by the stress ratio within the stable growth of fatigue crack(Region II) with an increase in $\Omega\textrm{K}$. The resistance to the fatigue crack growth at high temperature is higher in comparison with that at room temperature, and the resistance attributed to the extent of plasticity-induced by compressive residual stress according to the cyclic loads. Fractographic examinations reveal that the differences of the fatigue crack growth characteristics between room and high temperatures are mainly explained by the crack closure and oxide-induced by high temperature.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.534-545
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• 2015

In the very high temperature reactor (VHTR), which is a next generation nuclear reactor system, ceramics are used as a fuel coating material and graphite is used as a core structural material. Even if a depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change only slowly. This is because the thermal capacity of the core is so high. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel. The objectives of this study are to investigate the heat transfer characteristics of natural convection of a one-side heated vertical channel with inserted porous materials of high porosity and also to develop the passive cooling system for the VHTR. An experiment was carried out using a one-side heated vertical rectangular channel. To obtain the heat transfer and fluid flow characteristics of the vertical channel with inserted porous material, we have also carried out a numerical analysis using a commercial Computational Fluid Dynamics (CFD) code. This paper describes the thermal performances of the one-side heated vertical rectangular channel with an inserted copper wire of high porosity.

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