• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.1 s.173
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• pp.206-214
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tube s. the integrity evaluation must be carried out. and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire: integrity evaluation process. For this reason. an integrity evaluation system, which provides efficient of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). This system does not only provide various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications but also is compatible with other commercial database software. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.246-258
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• 2016

The shield building of AP1000 was designed to protect the steel containment vessel of the nuclear reactor. Therefore, the safety and integrity must be ensured during the plant life in any conditions such as an earthquake. The aim of this paper is to study the effect of water in the water tank on the response of the AP1000 shield building when subjected to three-dimensional seismic ground acceleration. The smoothed particle hydrodynamics method (SPH) and finite element method (FEM) coupling method is used to numerically simulate the fluid and structure interaction (FSI) between water in the water tank and the AP1000 shield building. Then the grid convergence of FEM and SPH for the AP1000 shield building is analyzed. Next the modal analysis of the AP1000 shield building with various water levels (WLs) in the water tank is taken. Meanwhile, the pressure due to sloshing and oscillation of the water in the gravity drain water tank is studied. The influences of the height of water in the water tank on the time history of acceleration of the AP1000 shield building are discussed, as well as the distributions of amplification, acceleration, displacement, and stresses of the AP1000 shield building. Research on the relationship between the WLs in the water tank and the response spectrums of the structure are also taken. The results show that the high WL in the water tank can limit the vibration of the AP1000 shield building and can more efficiently dissipate the kinetic energy of the AP1000 shield building by fluid-structure interaction.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.627-638
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• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.39 no.3
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• pp.369-380
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• 2019

When a extreme loading such as blast is applied to prestressed concrete (PSC) structures and infrastructures for an instantaneous time, serious property damages and human casualties occur. However, a existing design procedure for PSC structures such as prestressed containment vessel (PCCV) and gas storage tank do not consider a protective design for extreme internal blast scenario. Particularly, an internal blast is much more dangerous than that of external blast. Therefore, verification of the internal blast loading is required. In this paper, the internal blast resistance capacity of PSC member is evaluated by performing internal blast tests on RC and bi-directional PSC scaled down specimens. The applied internal blast loads were 22.68, 27.22, and 31.75 kg (50, 60, and 70 lbs) ANFO explosive charge at 1,000 mm standoff distance. The data acquisitions include blast pressure, deflection, strain, crack patterns, and prestressing force. The test results showed that it is possible to predict the damage area to the structure when internal blast loading occurs in PCCV structures.

• Magazine of the Korean Society of Agricultural Engineers
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• v.39 no.4
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• pp.106-113
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• 1997

Dynamic loading of structures often causes excursions of stresses well into the inelastic range, and the influence of the geometric changes on the dynamic response is also significant in many cases. Therefore, both material and geometric nonlinearity effects should be considered in case that a dynamic load acts on the structure. A structure in a nuclear power plant is a structure of importance which puts emphasis on safety. A nuclear container is a pressure vessel subject to internal pressure and this structure is constructed by a reinforced concrete or a pre-stressed concrete. In this study, the material nonlinearity effect on the dynamic response is formulated by the elasto-viscoplastic model highly corresponding to the real behavior of the material. Also, the geometrically nonlinear behavior is taken into account using a total Lagrangian coordinate system, and the equilibrium equation of motion is numerically solved by a central difference scheme. The constitutive relation of concrete is modeled according to a Drucker-Prager yield criterion in compression. The reinforcing bars are modeled by a smeared layer at the location of reinforcements, and the steel layer model under Von Mises yield criteria is adopted to represent an elastic-plastic behavior. To investigate the dynamic response of a nuclear reinforced concrete containment structure, the steel-ratios of 0, 3, 5 and 10 percent, are considered. The results obtained from the analysis of an example were summarized as follows 1. As the steel-ratio increases, the amplitude and the period of the vertical displacements in apex of dome decreased. The Dynamic Magnification Factor(DMF) was some larger than that of the structure without steel. However, the regular trend was not found in the values of DMF. 2. The dynamic response of the vertical displacement and the radial displacement in the dome-wall junction were shown that the period of displacement in initial step decreased with the steel-ratio increases. Especially, the effect of the steel on the dynamic response of radial displacement disapeared almost. The values of DMF were 1.94, 2.5, 2.62 and 2.66, and the values increased with the steel-ratio. 3. The characteristics of the dynamic response of radial displacement in the mid-wall were similar to that of dome-wall junction. The values of DMF were 1.91, 2.11, 2.13 and 2.18, and the values increased with the steel-ratio. 4. The amplitude and the period of the hoop-stresses in the dome, the dome-wall junction, and the mid-wall were shown the decreased trend with the steel-ratio. The values of DMF were some larger than those of the structure without steel. However, the regular trend was not found in the values of DMF.

• Journal of the Korea Concrete Institute
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• v.28 no.6
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• pp.673-683
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• 2016

In recent years, frequent terror or military attack by explosion, impact, fire accidents have occurred. Particularly, World Trade Center collapse and US Department of Defense Pentagon attack on Sept. 11 of 2001. Also, nuclear power plant incident on Mar. 11 of 2011. These attacks and incidents were raised public concerns and anxiety of potential terrorist attacks on major infrastructures and structures. Therefore, the extreme loading researches were performed of prestressed concrete (PSC) member, which widely used for nuclear containment vessel and gas tank. In this paper, to evaluate the blast resistance capacity and its protective performance of bi-directional unbonded prestressed concrete member, blast tests were carried out on $1,400{\times}1,000{\times}300mm$ for reinforced concrete (RC), prestressed concrete without rebar (PSC), prestressed concrete with rebar (PSRC) specimens. The applied blast load was generated by the detonation of 55 lbs ANFO explosive charge at 1.0 m standoff distance. The data acquisitions not only included blast waves of incident pressure, reflected pressure, and impulse, but also included displacement, acceleration, and strains at steel, concrete, PS tendon. The results can be used as basic research references for related research areas, which include protective design and blast simulation under blast loading.