• Proceedings of the KWS Conference
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• 2002.10a
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• pp.38-42
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• 2002

HAZ(Heat Affected Zone of weldm ents) properties were investigated for a high nitrogen austenitic stainless steel with a chemical composition of Fe-0.02C-0.15Si-6.00Mn-10.0Ni-23.0Cr-2.00Mo-0.48N-0.14V. Thermal cycle of HAZ was simulated by the thermal cycle simulator (Gleeble 1500). The heat treatment was applied to the Charpy test size sample without notch under various peak temperatures and/or the holding times condition. V-notch Charpy test was performed at the temperature range of 273~77 K. Metallographic examination also was carried out by using optical microscopy, scanning electron microscopy and transmission electron microscopy. The simulated specimens revealed a slight embrittlement compared with the base materials. The impact toughness of the specimens deteriorated with the decreasing test temperature. The results from Charpy V-notch test, however, showed that significant degradation of absorbed energy caused by brittle fracture was not observed for the specimen tested in the test temperature range.

• Korean Journal of Materials Research
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• v.7 no.7
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• pp.623-631
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• 1997

본 연구에서는 주조 2상 스테인리스강의열시효에 대한 시효온도, 시효시간 및 Nb함유량의 영향을 관찰하기 위해 기계적 성질 및 조직을 조사하였으며 Nb을 함유한 주조 2상 스테인리스강의 파괴기구를 규명하기 위해 SEM에 의한 파단면 관찰과 WDS성분분석을 통해 파괴기구의 특성을 고찰하였다. 시효온도와 시효시간이 증가함에 따라 페라니트으 미소경도가 증가하였으며 항복강도의 경우 시효온도와 시효시간에는 영향을 받지 않았으나 Nb을 함유한 재료들이 Nb을 함유치 않은 재료들에 비해 다소 낮은 항복강도 값을 보였다. 충격흡수에너지 값은 시효시간 및 시효온도의 증가에 따라 시험된 모든 재료에서 저하되었는데 0.4% Nb을 함유하는 경우 Nb을 약간 함유하거나 함유치 않은 재료들에 비해 시효시간에 따라 급격한 감소 경향을 보였다. 파단면 관찰결과 페라이트 기지 또는 페라이트/오스테나이트 상경계에서 석출된 VbC를 비롯한 탄화물들이 취성저항성을 낮추는데 크게 기여했음을 알 수 있었다.

• Journal of Welding and Joining
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• v.1 no.1
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• pp.47-55
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• 1983

The mechanical and microstructural properties in high heat input welds of home-made SM 50 high strength steels were investigated and compared with the manual shielded metal arc welds. Also, the fracture toughnesses of the simulated weld-bonds with various thermal cycles were quantatively examined in order to provide the basic data for further development of the high strength steels for high input welding. Main results obtained are as follows. (1) The embrittlement degree and the coarse grained region in high heat input welds appear to be extraordinarily large compared with the manual shielded metal arc welds, while the difference in change of nicrohardness is not so large in both welds. (2) The embrittleness in high heat input weld-bonds is mainly affected by the size of coarse grain rather than the microstructure. (3) The fracture toughness in high heat input weld-bonds can be improved by controlling the cooling rate from 800.deg.C to 500.deg.C rapidly.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.311-322
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• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2803-2811
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• 2020

With the aging of nuclear reactors, embrittlement of the reactor pressure vessel (RPV) steel, as a consequence of routine operations, is highly probable. To ensure operational integrity and safety, prediction and mitigation of compromising damage, brought on by pressurized thermal shock (PTS) following an emergency procedure, is of utmost importance. Computational fluid dynamics (CFD) codes can be employed to predict these events and have therefore been an acceptable method for such assessments. In this paper, CFD simulations of a density driven ECC state in the ROCOM facility are analyzed. Obtained numerical results are validated with the experimental measurements. Considerable attention is attributed to the boundary conditions and their influence, specifically outlet definitions, in order to determine and adequately replicate the non-active pumps in the facility. Consequent analyses focused on initial conditions as well as on the temporal discretization and inner iterations. Disparities due to different turbulent modelling approaches are investigated for standard RANS models. Based on observed trends for different cases, a definitive simulation setup has been established, results of which have been ultimately compared to the measurements.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.7
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• pp.1139-1146
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• 2001

The reactor pressure vessel(RPV) is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and a number of subclad cracks have been found during an in-service-inspection. These subclad cracks should be assured for a safe operation under normal conditions and faulted conditions such as pressurized thermal shock(PTS). Currently available integrity assessment procedure for an RPV, ASME Code Sec. XI, are built on the basis of linear fracture mechanics (LEFM). In PTS condition, however, thermal stress and mechanical stress give rise to high tensile stress at the cladding and elastic-plastic behavior is expected in this area. Therfore, ASME Code Sec. XI is overly conservative in assessing the structural integrity under PTS condition. In this paper, the fracture parameter (stress intensity factor, K, and RT(sub)NDT) from elastic analysis using ASME Sec. XI and finite element method were validated against 3-D elastic-plastic finite element analyses. The difference between elastic and elastic-plastic analysis became significant with increasing crack depth. Therfore, it is recommended to perform elastic-plastic analysis for the accurate assessment of subclad cracks under TPS which causes plastic deformation at the cladding.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• 한국신재생에너지학회:학술대회논문집
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• 2010.11a
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• pp.114.1-114.1
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• 2010

A highly performed Pd-based hydrogen membrane has prepared successfully on a modified porous nickel support. The porous nickel support modified by impregnation method of $Al(NO_3)_3{\cdot}9H_2O$ (Aldrich Co.) over the nickel powder showed a strong resistance to hydrogen embrittlement and thermal stability. Plasma surface modification treatment was introduced as a pre-treatment process instead of conventional HCl wet activation. Ceramic barrier was coated on the external surface of the prepared nickel supports to prevent intermetallic diffusion and to enhance the affinity between the support and membrane. Palladium and copper were deposited at thicknesses of $4\mu}m$ and $0.5{\mu}m$, respectively, on a barrier-coated support by DC sputtering process. The permeation measurement was performed in pure hydrogen at $400^{\circ}C$. The single gas permeation of our membrane was two times higher than that of the previous membrane which do not have ceramic barrier.

• Korean Journal of Materials Research
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• v.15 no.2
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• pp.126-133
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• 2005

Mechanical properties of zirconium cladding were evaluated by ring tension test to simulate Reactivity-Initiated Accident (RIA) at high burnup situation as an out-reactor test. Zircaloy-4 cladding was hydrided up to 1000 ppm as well as oxidized up to $100\;{\mu}m$ to simulate high-burnup situation. After simulated high-burnup treatment, ring tension test was carried out from 0.01 to 1/sec to correlate with actual RIA event. The results showed that ductility and circumferential toughness decreased with the hydrogen content and oxide thickness. Hydride generated inside cladding acted as brittle failure. Oxygen influenced cladding tube by the reduction of load bearing area, oxygen embrittlement, and thermal aging. Correlation between in-reactor RIA parameter like fuel enthalpy and out-reactor toughness was performed and showed a reasonable result.

• Transactions of the Korean hydrogen and new energy society
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• v.22 no.4
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• pp.458-464
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• 2011

Recently, one of the hydrogen production methods has attracted using dense metallic membrane. It has high hydrogen permeation and selectivity which hardly could adopt industrial product because of high cost, hydrogen embrittlment and thermal stability. Meanwhile, vanadium has high hydrogen solubility and it use to instead of Pd-Ag amorphous membrane. Aluminum carried out blocking hydrogen diffusion on grain boundary therefore protecting hydrogen embrittlement. Most of dense metallic membrane is solution diffusion mechanism. The solution diffusion mechanism was very similar hydrogen storing steps such as steps of metal hydride. Thus, V-Al composites were fabricated to use hydrogen induced mechanical alloying. The fabricated V-Al composites were characterized by XRD, SEM, EDS and simultaneous TG/DSC analyses. The hydrogenation behaviors were evaluated using a Sievert's type automatic PCT apparatus. The hydrogenation behaviors of V-Al composites was evaluated too low hydrogen stored capacity and fast hydrogenation kinetics. In PCI results, V-Al composites had low hydrogen solubility, in spite of that, hydrogen kinetics was calculated very fast and hydrogen absorption/desorption contents were same capacity.