• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.10-17
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• 2020

Securing the integrity of steam generator tubes is an essential requirement for safe operation of nuclear power plants. Therefore, tubes that do not satisfy integrity requirements are no longer usable and must be repaired according to the related requirements. In general, the repair criterion is that the damage depth is more than 40% of the tube wall thickness. However, the plugging limit can be changed and be applied, provided a technical proof is given that integrity can be secured against specific degradation at a specific plants and that approval can be obtained from a regulatory agency. A typical example is alternative repair criteria for defects within the tube sheet or tube support plates. In this paper, a background of establishing the plugging limit for steam generator tubes and changes in maintenance criteria are reviewed as examples.

• Journal of the Korean Society for Precision Engineering
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• v.32 no.11
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• pp.997-1005
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• 2015

In this study, modal analysis and equivalent static load analysis for valve supports of 26" gas piping in gas stations were conducted and the existing straight and inclined types of valve supports were compared using seismic performance testing. Also, a new valve support shape was suggested by optimizing position of fastener holes, width and thickness of the support, and size of bracket. Improvement in seismic performance by design optimization was verified through equivalent static load analysis. The seismic performance of the newly proposed valve support was greatly improved and the maximum displacement and maximum stress of the seismic load was about 20% lower than those of the existing valve support.

• Tunnel and Underground Space
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• v.13 no.1
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• pp.1-5
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• 2003

The United States Antarctic program (USAP) through its principal Support Contractor Raytheon polar Services Co. (RPSC), has recently finished a 3 years projects, almost 936m length of underground utility tunnels at Amundsen-Scott station. It accommodates the piping that conveys fresh water from current well sites, as well as waste water to repositories in abandoned wells. The under snow tunnels allow year-round access for system operations and maintenance.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1323-1332
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• 2009

The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2007.04a
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• pp.322-325
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• 2007

This paper shows the development procedure of the low pressure LOX feed system which is used in the liquid rocket with a turbopump. Korean Air has cooperated with KARI in developing the LOX feed system to turbopump. The LOX feed system is characterized with cryogenic temperature and the thin-thickness tube for weight saving. The system in this project is composed with a main feed line and a recirculation line for the LOX temperature conditioning. Each piping system has many components, namely, bellows, filter, orifice, valves, flange and support. In this paper, system design & manufacturing, structural & thermal analyses, and component tests are explained. Finally, the system was assembled to the KARI's PTF test facility and functioned well to meet its required performance.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.2
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• pp.7-13
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• 2011

Turbine journal bearings are designed to support the weight of the rotors on a hydrodynamic oil film and to provide dynamic stability to the rotor system. The life time of journal bearings is infinite theoretically because the journal bearings are separated from the shaft journal by oil film. But poor design, assembly, operation and maintenance can cause problems to the journal bearings. The FMEA(Failure Mode and Effects Analysis) results of the journal bearings show that frequent maintenance of the journal bearings can cause failures and reduction of the bearing life. Therefore, the maintenance periods and history of the journal bearings with the bearing FMEA results are reviewed in order to establish the optimized maintenance period of the journal bearing for the nuclear power plants. Consequently it is necessary to maintain a best condition of lubrication system, reject time-based maintenance and perform the condition-based maintenance of journal bearings in order to maintain optimum condition of the journal bearing.

• 유체기계공업학회:학술대회논문집
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• 2002.12a
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• pp.419-424
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• 2002

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality In February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. A flow simulation facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The flow circulation system is composed of a circulation pump, a core flow pipe, a core bypass flow pipe and instruments. The system is to be filled with de-mineralized water and the flow should be met the design flow to simulate similar flow characteristics in the core channel of the half-core test facility to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the system. The computational flow analysis has been performed for the verification of system pressure variation through the three-dimensional analysis program with standard k-$\epsilon$ turbulence model and for the verification of the structural piping integrity through the finite element method. The results of the analysis are satisfied the design requirements and structural piping integrity of flow circulation system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.20-24
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• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.11-16
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• 2016

Fuel assembly's mechanical characterization test facility (FAMeCT) in KAERI was constructed with upgraded functional features such as increased loading capacity, underwater vibration testing and severe earthquake simulation for extended fuel design guideline. This facility is designed and developed to provide out-pile fuel data for accident analysis model and fuel licensing. Functional tests of FAMeCT were performed to confirm functionality, structural integrity, and validity of newly-built fuel assembly mechanical test facility. Test program includes signal check of data acquisition system, load delivering capacity using real-sized fuel assemblies and a standard loading cylindrical rigid specimen. Fuel assembly's lateral bending test was carried out up to 30 mm of pull-out displacement. Limit case axial compression loading test up to 33 kN was performed to check structural integrity of UCPS (Upper Core Plate Simulator) support frame. Test results show that all test equipment and measurement system have acceptable range of alignment, signal to noise ratio, load carrying capacity limit without loss of integrity. This paper introduces newly constructed fuel assembly's mechanical test facility and summarizes results of functional test for the mechanical test equipment and data acquisition system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.30-36
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• 2020

In SMART, the flow mixing header assembly (FMHA) is used to mix the coolant flowing into the reactor core to maintain a uniform temperature. The FMHA is designed to have enough stiffness so the resonance with reactor internal structures does not occurs during the pipe break and the seismic accidents. Since the gap between the FMHA and the core support barrel assembly is very narrow compared with the diameter of FMHA, the hydrodynamic mass effect acting on the FMHA is not negligible. Therefore the hydrodynamic mass characteristics on the FMHA are investigated to consider the fluid and structure interaction effects. The result of modal analysis for the dry and underwater conditions, the natural frequency of primary vibration mode for the horizontal direction is reduced from 136.67 Hz to 43.76 Hz. Also the result of frequency response spectrum seismic analysis for the dry and underwater conditions, the maximum equivalent stress are increased from 13.89 MPa to 40.23 MPa. Therefore, reactor internal structures located in underwater condition shall consider carefully the hydrodynamic mass effects even though they have sufficient stiffness required for performing its functions under the dry condition.