• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.54-60
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• 2011

A spacer grid, which supports nuclear fuel rods laterally and vertically with a friction grip, is one of the most important structural components in a PWR fuel. The form of grid strap and supporting parts such as grid spring and dimple is known to be closely related with the mechanical/structural performance of spacer grid and nuclear fuel assembly. In this study, reviewing various research results for enhancing the performance of the spacer grid, some structural design considerations and research directions on the spacer grid assembly are suggested for further study.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.1-8
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• 2010

Nuclear steam generator tubes have experienced wear degradation at tube support structure. Morphology of wear scar was analyzed by using eddy current signal. A burst test facility for steam generator tubes was established and tubes with 3 types of defects were tested. The burst test results show that the depth of wear scar is the main factor influencing the burst pressure of tubes, meanwhile, both the longitudinal length and the angle also have effect on the burst pressure. Based on test results, the burst pressure equation for wear degradation was proposed.

• 한국압력기기공학회 논문집
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• 제16권2호
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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.

• 한국정밀공학회지
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• 제32권11호
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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.

• 터널과지하공간
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• 제13권1호
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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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• 제41권10호
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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.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2007년도 제28회 춘계학술대회논문집
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• pp.322-325
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• 2007

본 논문은 터보펌프를 사용하는 액체로켓의 저압 LOX 공급계의 개발 프로세스를 제공한다. 대한항공은 한국항공우주연구원과 협력하여 터보펌프 공급을 위한 LOX 공급계 개발을 수행하였다. LOX 공급계는 극저온의 온도와 무게절감을 위한 얇은 배관두께가 특징이다. 본 프로젝트의 시스템은 주 배관과 LOX 온도 제어를 위한 재순환 배관으로 구성되어 있다. 각 배관시스템은 벨로우즈, 필터, 오리피스, 밸브류, 플랜지와 서포트로 구성되어 있다. 이 논문에서는 시스템 설계 및 제작, 구조 및 열 해석, 단품 시험에 대하여 설명하였다. 최종적으로, 이 시스템은 한국항공우주연구원의 PTF 시험설비에 조립되어 요구 성능을 달성하였다.

• 한국압력기기공학회 논문집
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• 제7권2호
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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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• 유체기계공업학회 2002년도 유체기계 연구개발 발표회 논문집
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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.

• 한국압력기기공학회 논문집
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• 제9권1호
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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.