• Journal of Mechanical Science and Technology
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• 제20권12호
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• pp.2061-2078
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• 2006

The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500$^{\circ}C$ and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1115-1126
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• 2009

As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.648-653
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• 2003

In order to predict stress intensity limit of high-temperature creep structures, creep work-time equation, defined as $W_ct^P=B$, was used, and the results of the equation were compared with isochronous stress-strain curve (ISSC) ones of ASME BPV NH Code. For this purpose, the creep strain tests with. time variations for commercial type 316 stainless steel were conducted with different stresses; 160 MPa, 150 MPa, 145 MPa, 140 MPa and 135 MPa at $593^{\circ}C$. The results of log $W_c$ and log t plots showed a good linear relation up to $10^5$ hr. The constants p, B and stress intensity limit values showed comparatively good agreement to those of ASME NH ISSC. It is believed that the relation can be simply obtained with only several short-term 1% strain data without ISSC which can be obtained by long-term creep data.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.101-102
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• 2005

• 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.

• 대한기계학회논문집A
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• 제38권8호
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• pp.831-836
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• 2014

본 연구에서는 한국원자력연구원 내에 설치될 예정인 소듐시험 시설인 SELFA(Sodium Thermal-hydraulic Experiment Loop for Finned-tube Sodium-to-Air heat exchanger) 내에서 정상상태 가동온도가 $510^{\circ}C$의 고온 압력용기인 팽창탱크에 대해 고온 건전성 평가를 수행하였다. 팽창탱크에 대해 3 차원 유한요소해석에 기초하여 고온설계 기술기준인 ASME Section III Subsection NH 와 프랑스의 RCC-MRx 코드를 따라 크리프-피로 손상평가를 수행하였다. 평가결과 팽창탱크는 크리프-피로 설계 과도 하중 하에서 구조적 건전성을 유지하는 것으로 나타났다. 316L 스테인리스강 재질의 동 압력용기에 대해 정량적 코드 비교 분석을 수행하였다.

• 대한기계학회논문집A
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• 제37권10호
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• pp.1251-1259
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• 2013

본 연구에서는 소듐냉각 고속로 붕괴열교환기(DHX)의 고온 설계 및 크리프-피로 손상 평가를 수행하였다. 제 4 세대 소듐냉각 고속로의 능동 및 피동 잔열제거계통에 설치되는 DHX와 한국원자력연구원의 STELLA-1 시험루프에 설치된 DHX에 대해 상세설계 및 3D 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준인 ASME Section III Subsection NH와 RCC-MR 코드를 따라 크리프-피로 손상평가를 수행하였다. 크리프-피로 손상평가 결과에 기초하여 두 설계기준에 대해 비교 분석하고, 설계 기술기준의 보수성 이슈에 대해 토의하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.823-830
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• 1998

본 연구에서는 미국, 프랑스, 일본의 고온구조 설계지침서의 크립-피로 손상평가 방법을 살펴보고 고온하중을 받는 불연속 구조물에 대하여 범용 유한요소 해석코드인 ANSYS와 ABAQUS를 이용한 열전달 및 응력해석을 수행하여 각국의 코드 절차에 따른 크립-피로 손상 평가를 하였다. 이들 결과를 점소성 비탄성 구성식을 적용한 비탄성해석 결과와 비교평가하였다. 본 연구에서 분석한 불연속 구조물의 경우에 대한 평가 결과 미국의 ASME Subsection NH에 의한 방법이 비탄성 해석결과에 가장 가까운 결과를 주며, 일본의 BDS에 따른 평가방법은 적용성이 가장 편리함을 알 수 있었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.837-843
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• 1998

This paper presents the thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the secondary piping of LMFR using Green's function method and standard FEM. The thermohydraulic loading conditions used in the present analysis are simplified sinusoidal thermal loads and the random type data thermal load. The thermomechainical fatigue damage was evaluated according to ASME code subsectionNH. The analysis results of fatigue for the sinusoidal and random load cases show that fatigue failure would occur at a geometrically discontinuous location during 90,000 hours of operation The fracture mechanics analysis showed that the crack would be initiated at an early stage of the operation. The fatigue crack was evaluated to propagate up to 5 ㎜ along the thickness direction during the first 944 and 1083 hours of operation for the sinusoidal and the random loading cases, respectively.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 1999년도 추계 학술발표회 논문집 Proceedings of EESK Conference-Fall
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• pp.220-227
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• 1999

The Purpose of this paper is to evaluate the buckling strength of conceptually designed KALIMER reactor vessel. For evaluation of the buckling load buckling load the design equations and the finite element analysis are used. In finite element method the eigenvalue buckling analysis nonlinear elastic buckling analysis using snap-through buckling method and nonlinear elastic-plastic buckling analysis are carried out. the calculated buckling loads of KALIMER reactor vessel using the finite element method are in well agreement with those of the design equations. From the calculated results of buckling load in KALIMER rector vessel it is shown that the plasticity of vessel materials significantly affects the buckling load but the initial imperfection has little effects, In checking the limits of bucking load of KALIMER reactor vessel using the ASME B & PV Section III. Subsection NH the non-seismic isolation design can not satisfy the buckling limit requirements but the seismic isolation design can sufficiently satisfy the requirements.