• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.61-68
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• 2017

Reactor pressure vessel (RPV) under severe accident conditions accompanied by core melting is exposed to direct high-temperature thermal loads. Understanding the creep behavior of the material is one of the most important factors for evaluating the structural integrity at these conditions. While damage evaluation studies have been conducted on critical structures of nuclear power plants through finite element (FE) analyses considering creep behavior, for accurate creep damage evaluation, constitutive equations considered in the FE analyses may have different results depending on the time hardening and strain hardening models as well as the tertiary creep consideration. The purpose of this study is to evaluate the creep damage under severe accident conditions by using FE method for a representative domestic RPV material, SA508 Gr.3. The effect of material hardening models and constitutive equations which are the main variables were also investigated.

• 전기의세계
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• v.19 no.4
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• pp.18-23
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• 1970

A neutron detector, in general, can not be utilized as the thermal neutron detecting chamber in the nuclear power reactor, especially P.W.R. due to the characteristics of high temperature, high pressure and high neutron flux in a reactor vessel. We have performed an experiment to detect the thermal neutrons at 400.deg. C and high flux of thermal neutron in a power reactor. Coating boron-10 on the aluminium plates by means of surface diffusion method at 600.deg. C for 5 hours in an electric furace, also we made a typical chamber which was compensated ionization chamber filled with free air as an ionization gas. It was checked the chamber characteristics in the TRIGA MARK-II Reactor at the power level from zero to 250KW. The chamber current showed a perfect linear increase to power increase. However, many variation of the measured current were observed within the power of 50KW.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1338-1348
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• 2016

A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at $500^{\circ}C$ to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.4
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• pp.756-763
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• 2001

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 evaluation 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, tow dimensional finite element analyses were applied for various surface cracks. A total of 18 crack geometries were analyzed, and $\Omega$ 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-tip stress field due to constraint effect.

• Journal of the Korean Institute of Gas
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• v.2 no.4
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• pp.1-6
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• 1998

In order to increase design efficiency, it is required to design steam boiler pressure parts systematically considering daily start-stop operation and load variation. The objective of this research is to develope an integrated system for increasing design efficiency of boiler pressure parts. The developed system consists of three program modules: (1) flexibility design module for the header stub considering fatigue life, (2) fatigue limit calculation and life evaluation module for the thick-walled boiler pressure part under cyclic operation using TRD301 code, (3) drawing automation module for the header and drum producing design drawings, welding data and bill of materials.

• Journal of Mechanical Science and Technology
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• v.18 no.12
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• pp.2310-2318
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• 2004

The effects of change in injection pressure on spray structure in high temperature and pressure field have been investigated. The analysis of liquid and vapor phases of injected fuel is important for emissions control of diesel engines. Therefore, this work examines the evaporating spray structure using a constant volume vessel. The injection pressure is selected as the experimental parameter, is changed from 400 bar to 800 bar by using a common rail injection system. Also, we conducted simulation study by modified KIVA-II code. The results of simulation study are compared with experimental results. The images of liquid and vapor phase for free spray were simultaneously taken by exciplex fluorescence method. As experimental results, the vapor concentration of injected fuel is leaner due to the increase of atomization in the case of the high injection pressure than in that of the low injection pressure. The calculated results obtained by modified KIVA-II code show good agreements with experimental results.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.17 no.4
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• pp.23-31
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• 2018

Stress of a steam separator, equipment of the high-pressure (HP) evaporator for a HRSG, was analyzed and evaluated according to ASME Boiler & Pressure Vessel Code Section VIII Division 2. First, from the analysis results of the piping system model of the HP evaporator, reaction forces of the riser tubes connected to the steam separator, i.e., nozzle loads, were derived. Next, a finite element model of the steam separator was constructed and analyzed for the design pressure and the nozzle loads. The results show that the maximum stress occurred at the bore of the riser nozzle. The primary membrane stresses at the shell and nozzle were found to be less than the allowable stress. Next, the steam separator was analyzed for the steady-state operating conditions of operating pressure, operating temperature, and nozzle loads. The maximum stress occurred at the bore of the riser nozzle. The primary plus secondary membrane plus bending stress at the shell and nozzle was found to be less than the allowable stress.

• Computers and Concrete
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• v.14 no.2
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• pp.127-143
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• 2014

LOCA (Loss Of Coolant Accident) is one of the most important utmost accidents for Prestressed Concrete Containment Vessel (PCCV) due to its coupled effect of high temperature and inner pressure. In this paper, heat conduction analysis is used to obtain the LOCA temperature distribution of PCCV. Then the elastic internal force of PCCV under LOCA temperature is analyzed by using both simplified theoretical method and FEM (finite element methods) method. Considering the coupled effect of LOCA temperature, a nonlinear elasto-plasitic analysis is conducted for PCCV under utmost internal pressure considering three failure criteria. Results show that the LOCA temperature distribution is strongly nonlinear along the shell thickness at the early time; the moment result of simplified analysis is well coincident with the one of numerical analysis at weak constraint area; while in the strong constrained area, the value of moments and membrane forces fluctuate dramatically; the simplified and numerical analysis both show that the maximum moment occurs at 6hrs after LOCA.; the strain of PCCV under LOCA temperature is larger than the one of no temperature under elasto-plastic analysis; the LOCA temperature of 6hrs has the greatest influence on the ultimate bearing capacity with 8.43% decrease for failure criteria 1 and 2.65% decrease for failure criteria 3.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.344-357
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• 2002

The present paper reports spontaneous steam explosions observed in fuel coolant interaction experiments using prototypic reactor materials. Pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$ are used. A high temperature molten material in the form of a jet is poured into a subcooled water pool located in a pressure vessel. An induction skull melting technique is used for the melting of the reactor material. In both tests using pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$, either a quenching or a spontaneous steam explosion was observed. The morphology of debris and pressure profile clearly indicate the differences between the qunching cases and explosion cases. The dynamic pressure. dynamic impulse, water temperature, melt temperature, and static pressure Inside the containment chamber were measured . As the spontaneous steam explosion for the reactor material is firstly observed in the present experiments, the results of present experiments could be a siginificant step forward the understanding the explosion of the reactor material.

• Journal of the Korean Society of Safety
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• v.32 no.2
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• pp.1-6
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• 2017

Modern society has been experiencing by population growth and urbanization that bring, a change of eating habits which has occurred a various types of waste in a large amount. Even though these wastes are required an immediate treatment with difficulties unsanitary handling and existing waste treatment method are by incineration, fermentation, drying and etc. however a bad smell occurs after the treatment that need's a lot of energy in processing organic wastes with high moisture contents and wasteful and inefficient problem. The strength assessment of the organic waste agitating vessel is required in terms of safety due to the differences of loading on the shaft that was treated by agitating the mixture of food waste. The damage of agitating axis is depended on steam pressure, temperature condition and the force moment that exerted by the food waste. Thus the strength assessment and stability evaluation are very important, especially to handle a hard waste. In this study the rotation capacity of agitation is about 5 tons considering general structural rolled steel pressure vessel strength and steam pressure. The purpose is to estimate the safety and strength evaluation for a agitator axis and impellers according to the rotating angle of the axis under the condition of the 3.2 ton capacity reactor.