• Nuclear Engineering and Technology
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• 제38권3호
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• pp.293-300
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• 2006

A liquid metal reactor (LMR) operated at high temperatures is subjected to both cyclic mechanical loading and thermal loading; thus, creep-fatigue is a major concern to be addressed with regard to maintaining structural integrity. The Korea Advanced Liquid Metal Reactor (KALIMER), which has a normal operating temperature of $545^{\circ}C$ and a total service life time of 60 years, is composed of various cylindrical structures, such as the reactor vessel and the reactor baffle. This study focuses on the creepfatigue crack initiation for a cylindrical Y-junction structure made of 316 stainless steel (SS), which is subjected to cyclic axial tensile loading and thermal loading at a high-temperature hold time of $545^{\circ}C$. The evaluation of the considered creep-fatigue crack initiation was carried out utilizing the ${\sigma}_d$ approach of the RCC-MR A16 guide, which is the high-temperature defect assessment procedure. This procedure is based on the total accumulated strain during the service time. To confirm the evaluated result, a high-temperature creep-fatigue structural test was performed. The test model had a circumferential through wall defect at the center of the model. The defect front of the test model was investigated after the $100^{th}$ cycle of the testing by utilizing a metallurgical inspection technique with an optical microscope, after which the test result was compared with the evaluation result. This study shows how creep-fatigue crack initiation for a high-temperature structure can be predicted with conservatism per the RCC-MR A16 guide.

• 비파괴검사학회지
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• 제20권5호
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• pp.390-396
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• 2000

원자로 재료인 SA 508 Class 3 강용접부 및 열영향부 모사 시험편에 대해서 초음파공명분광법으로 동적탄성계수를 측정하였다. 등방성 탄성계수를 가정하여 초기 추정 탄성 계수, $c_{11},\;c_{12}$ 및 $c_{44}$로부터 장방형 시편의 공명 주파수를 계산하였으며 계산된 주파수와 초음파공명분광법으로 측정된 주파수를 비교, 반복 수렴 절차를 거쳐 정밀한 탄성계수를 구했다. 열처리 조건의 차이 및 미세 조직의 차이에 따라 영률 및 전단 계수의 차이가 확실하게 나타났다. 미세한 베이나이트 조직에서의 영률 및 전단 계수는 조대한 마르텐사이트 조직보다 높았으며 이러한 경향은 미세 경도 시험 등의 다른 실험 결과와도 일치하였다.

• 대한기계학회논문집
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• 제14권1호
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• pp.112-118
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• 1990

본 연구에서는 ASTM-E24.01.06에서 제안하고 있는 실험방법을 응용하여 균열 정지 파괴인성값을 측정하였다.즉 쐐기와 분리형 부싱(wedge and split bushing)으 로 압축하중을 가함으로 균열선 웨지하중 시편[crack line wedge loaded specimen(CL- WL시편)]에 인장력을 발생시켜서 균열정지 응력확대계수( $K_{1a}$)를 결정하였다. 그리고 균열개시 응력확대계수가 균열정지 응력확대계수에 미치는 영향들을 여러가지 재료들에 대하여 체계적으로 검토하였다.다.

• 대한기계학회논문집A
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• 제36권5호
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• pp.513-521
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• 2012

본 연구의 목적은 손상모델의 온도의존성을 고려하여 압력용기강의 취성파괴 거동을 평가하는 것이며, 이를 위해 다중 섬 유전자알고리즘과 와이블 응력모델을 연계하여 대표적 취성파괴 손상모델의 재료상수 결정절차를 개선하였다. 벽개파괴가 예상되는 $-60^{\circ}C$, $-80^{\circ}C$, $-100^{\circ}C$ 온도에서 수행한 SA508 탄소강 재료의 파괴 인성 실험 데이터를 사용하여 개선된 절차에 따른 재료상수를 결정하였고, NUREG/CR-6930과 동일한 결과인 재료상수의 온도의존성을 확인하였다. 최종적으로는 손상모델 재료상수의 온도의존성에 따른 2-매개변수 와이블 응력모델과 3-매개변수 와이블 응력모델의 차이를 정량화하였으며, 공학적으로 활용 가능한 관계식을 제안하였다.

• 대한금속재료학회지
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• 제47권3호
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• pp.139-146
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• 2009

The heat-affected zone (HAZ) of SA508 Gr.4N Ni-Mo-Cr low alloy steel, which has higher Ni and Cr contents than SA508 Gr.3 Mn-Mo-Ni low alloy steel, was investigated on the microstructure and mechanical properties. The HAZ was categorized into seven characteristic zones (CGCG, FGCG, ICCG, SCCG, FGFG, ICIC and SCSC-HAZ) according to the peak temperature from the thermal cycle experienced during multi-pass welding. Post Weld Heat Treatment (PWHT) was conducted in the temperature range of $550{\sim}610^{\circ}C$ for 30 hours to evaluate the effect of PWHT conditions on the microstructure and mechanical properties. Before PWHT, CGHAZ and FGFGHAZ showed high yield strength (YS) ranging from 1000 to 1250 MPa, while YS of SCSCHAZ decreased from 607 MPa (observed for base metal) to 501 MPa. The Charpy impact energies of sub-HAZs fell below 100J at $-29^{\circ}C$, except in the SCSCHAZ. By applying PWHT to sub-HAZ specimens, YS decreased as the PWHT temperature increased. In the case of CGHAZs and FGFGHAZ heat-treated at $610^{\circ}C$, YS dropped drastically to the range of 654~686 MPa. From the Charpy impact test, the upper-shelf energy (USE) increased to approximately 250J and Index temperature ($T_{68J}$) decreased below $-50^{\circ}C$. Specifically, in FGFG, ICIC and SCSC-HAZ, $T_{68J}$ was below -110, which was lower than the case of base metal.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1547-1554
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• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.340-350
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• 2015

This paper proposes strain-based acceptance criteria for assessing plastic instability of the safety class 1 nuclear components made of ferritic steel during level D service loads. The strain-based criteria were proposed with two approaches: (1) a section average approach and (2) a critical location approach. Both approaches were based on the damage initiation point corresponding to the maximum load-carrying capability point instead of the fracture point via tensile tests and finite element analysis (FEA) for the notched specimen under uni-axial tensile loading. The two proposed criteria were reviewed from the viewpoint of design practice and philosophy to select a more appropriate criterion. As a result of the review, it was found that the section average approach is more appropriate than the critical location approach from the viewpoint of design practice and philosophy. Finally, the criterion based on the section average approach was applied to a simplified reactor pressure vessel (RPV) outlet nozzle subject to SSE loads. The application shows that the strain-based acceptance criteria can consider cumulative damages caused by the sequential loads unlike the stress-based acceptance criteria and can reduce the overconservatism of the stress-based acceptance criteria, which often occurs for level D service loads.

• 한국안전학회지
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• 제30권3호
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• pp.107-113
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• 2015

The goal of this paper is to investigate structural integrity factors of RMI(reflective metal insulation) to confirm the design requirements in nuclear power plant. Currently, a glass wool insulation is using now, but it will gradually be replaced with the reflective metal insulation maded by stainless steel plates. The main function of an insulation is to minimize a heat loss of vessel and pipes in RCS(reactor coolant system). It has to maintain structural a integrity in nuclear power plant life duration. In this study, the structural integrity analysis was carried out both multi-plate and outer shell plate by using a static analysis and experimental test. First, inner multi-plate has a self support structure for being air space. Because the effect of total static weight in multi-layer plate is low, a plate collapse possibility is not high. Considering optimum thin plate pressing process, it has to pre-check the basic physical properties. Second, the outer segment thickness and stiffener shape are verified by the numerical static analysis, and sample test for both type of panel and cylindrical pipe model.

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1343-1354
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• 2022

Experiments were performed to measure the critical heat flux (CHF) on a vertical surface abutting a coarse packed bed of spherical particles. This geometry is representative of a CANDU reactor calandria tubesheet facing the end shield cavity during the in-vessel retention (IVR) phase of a severe accident. Deionized light water was used as the working fluid. Low carbon steel shielding balls with diameters ranging from 6.4 to 12.7 mm were used, allowing for the development of an empirical correlation of CHF as a function of shielding ball diameter. Previously published data is used to develop a more comprehensive empirical correlation accounting for the impacts of both shielding ball diameter and heating surface height. Tests using borosilicate shielding balls demonstrated that the dependence of CHF on shielding ball thermal conductivity is insignificant. The deposition of iron oxide particles transported from shielding balls to the heating surface is verified to increase CHF non-trivially. The results presented in this paper improve the state of the knowledge base permitting quantitative prediction of CHF in the CANDU end shield, refining our ability to assess the feasibility of IVR. The findings clarify the mechanisms governing CHF in this scenario, permitting identification of potential future research directions.

• 대한금속재료학회지
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• 제49권8호
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• pp.628-634
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• 2011

SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at $610^{\circ}C$ for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.