• 대한기계학회논문집A
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• 제33권2호
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

• 비파괴검사학회지
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• 제25권5호
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• pp.377-383
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• 2005

석유화학 및 정유설비는 고온이나 고압에서 폭발 위험성을 지닌 유체를 사용하기 때문에 방재기술에 관한 관심이 높다. 이들 설비 중에서도 고압반응기는 특히 고온/고압 하에서 사용되므로 안전성이 요구된다. 본 연구에서는 석유화학 플랜트의 고압반응기 소재로 많이 사용되고 있는 2.25Cr-1Mo강을 대상으로 하였으며, 3가지 온도조건에서 열화시간을 달리하여 총 8 종류의 인공열화 시험편을 준비하였다. 열화는 고압반응기의 사용온도인 $391^{\circ}C$보다 약간 높은 온도에서 둥온 열처리하였다. 미열화재를 포함하여 인공열화재에 대해 비커스경도값과 전기비저항값을 측정하였으며, Larson-Miller parameter와의 상관관계로부터 master curve를 작성하였다. 그리고 현장의 고압반응기에서 비커스경도와 전기비저항을 측정하여 실험실에서 작성한 master curve와 비교하였다. 전기비저항법을 이용한 고압반응기의 현장에서의 열화평가 가능성을 검토하였으며, 현장에서 측정한 전기비저항은 비슷한 열화수준에서의 인공열화재의 전기비저항값과 비슷하였다.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2407-2417
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• 2021

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

• 한국가스학회지
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• 제9권1호
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• pp.26-32
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• 2005

본 논문에서는 고압용기의 밀봉을 위해 사용하는 메탈시일의 설계 안전성에 대한 연구결과를 제시하고 있다. 기본적으로 고압용기의 1차적 밀봉은 메탈시일에 의해 이루어지고, 2차적 밀봉은 탄성체 고무로 제작된 0-링에 의해 가스누출이 차단된다. 유한요소해석 결과에 의하면, 가스누출 차단을 위해 사용한 알루미늄 소재의 밀봉장치는 강재를 사용한 경우보다 우수한 밀봉성을 보여주고 있다. 이것은 알루미늄 소재의 열팽창 특성이 상대적으로 우수하기 때문이다. 메탈시일과 압력용기 구조물에 걸리는 변형과 응력분포는 외부에서 공급된 가스압보다는 용기 내부에서 전달된 온도차에 의해 크게 지배를 받는 것으로 알려져 있다. 따라서 메탈시일의 소재는 통상적으로 $200^{\circ}C$이하로 유지되는 것이 소재의 강도와 밀봉성 측면에서 안전하다.

• 한국기계가공학회지
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• 제6권4호
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• pp.81-85
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• 2007

A safety valve is used for protecting the pressure vessel and facilities by discharging the operating fluid into the valve from the accident when the pressure is over the designated value. The fluid is sulfurous acid and nitric acid. etc. in the semi-conductor assembly line. Thus the valve elements material must be acid resistance. Teflon, which is used generally as inner parts of a valve, tends to easily sticks to sliding surface by thermal expansion under high temperature. Some studies are performed to change teflon to another material and shape to have a better fluidity under the condition. The analysis of the thermal expansion is conducted by commercial FEM software to improve the problems. Boundary conditions were temperature and load in this study. From the analysis, the thermal expansion of stainless steel is verified to be lower than that of teflon under high temperature. Thus coupled teflon/stainless steel-made valve is applied to assembly line without danger due to thermal expansion.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집B
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• pp.67-76
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• 2000

Cr-Mo low alloy steels have been used for a long time for pressure vessel due to its excellent corrosion resistance, high temperature strength and toughness. The paper reviewed the latest trends on material development and some problems on Cr-Mo low alloy steel for pressure vessel, such as elevated temperature strength, hardenability, synergetic effect between temper and hydrogen embrittlement, hydrogen attack and hydrogen induced disbonding of overlay weld-cladding.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.67-76
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• 2000

Cr-Mo low alloy steels have been used for a long time for pressure vessel due to its excellent corrosion resistance, high temperature strength and toughness. The paper reviewed the latest trends on material development and some problems on Cr-Mo low alloy steel for pressure vessel, such as elevated temperature strength, hardenability, synergetic effect between temper and hydrogen embrittlement, hydrogen attack and hydrogen induced disbonding of overlay weld-cladding.

• 한국산학기술학회논문지
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• 제21권2호
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• pp.66-73
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• 2020

현재 개발되고 있는 수소 압축 사이클에서는 압축기를 통해 초고압으로 압축된 수소를 고압용기 내에 저장하여 사용한다. 이러한 충전과정 중 용기내의 수소의 압력 및 온도 상승으로 인하여 고압용기에서 열응력이 발생할 수 있다. 고압용기의 신뢰성을 확보하기 위해서는 용기내의 수소의 온도를 예측하고 제어하는 것이 중요하다. 본 논문에서는 이러한 고압용기의 신뢰성 해석을 위하여 50 MPa급 수소압축시스템에서 고압용기를 충전하는 과정에서의 압력상승에 따른 용기 내의 수소온도 변화 및 외부와의 열평형까지 걸리는 시간, 감압밸브를 지날 때의 수소온도 변화, 고압용기 냉각을 위한 열교환기의 요구능력 등에 대하여 이론적인 방법과 수치적인 방법으로 해석을 수행하였다. 이론해석 결과, 고압용기의 내부 온도는 충전하기 전에 40 ℃에서 충전 후 1^st cycle, 2^nd cycle에서 평균적으로 126.675 ℃, 62.1 ℃가 증가하였다. 또한, 고압용기의 충전량은 1^st cycle, 2^nd cycle에서 각각 7.9 kg, 8.9 kg으로 계산되었다. 본 연구의 결과는 수소충전소와 같이 수소압축시스템이 필요한 현장의 인프라 설계 및 구축 등에 유용하게 활용 될 것이다.

• 한국기계가공학회지
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• 제17권4호
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• pp.150-159
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• 2018

The present research deals with a finite element analysis and fatigue evaluation of a steam separator of a high-pressure evaporator for the Heat Recovery Steam Generator (HRSG). The fatigue during the expected life of the HRSG was evaluated according to the ASME Boiler and Pressure Vessel Code Section VIII Division 2 (ASME Code). First, based on the eight transient operating conditions prescribed for the HRSG, temperature distribution of the steam separator was analyzed by a transient thermal analysis. Results of the thermal analysis were used as a thermal load for the structural analysis and used to determine the mean cycle temperature. Next, a structural analysis for the transient conditions was carried out with the thermal load, steam pressure, and nozzle load. The maximum stress location was found to be the riser nozzle bore, and hence fatigue was evaluated at that location, as per ASME Code. As a result, the cumulative usage factor was calculated as 0.00072 (much less than 1). In conclusion, the steam separator was found to be safe from fatigue failure during the expected life.

• 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.