• 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.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.134.1-134.1
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• 2005

• Journal of the Korean Society for Heat Treatment
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• v.25 no.3
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• pp.115-120
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• 2012

It has been known that the ferritic stainless steel can be changed to martensitic stainless steel when nitrogen is added. However the high hardness of martensitic stainless steel prevents the plastic deformation. In this study, instead of martensite, the surface microstructure was changed into nitrogen pearlite to increase the plastic deformation easily by isothermal heat treatment after high temperature gas nitriding (HTGN) the AISI 430 ferritic stainless steel. The isothermal treatment was carried out at $780^{\circ}C$ for 4, 6, and 10 hrs, respectively, after HTGN treatment at $1100^{\circ}C$ for 10 hrs. The surface layer of isothermal-treated steel appeared nitrogen pearlite composed with fine chromium nitride and ferrite. Hence, the interior region that was not affected by nitrogen permeation exhibited ferrite phase. When quenching the isothermal treated steel at 1100oC, martensitic phase formed at the surface layer. The hardness of surface layer of isothermal-treated steel and quenched steel measured the value of 150~240 Hv and 630 Hv, respectively.

• KEPCO Journal on Electric Power and Energy
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• v.1 no.1
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• pp.109-113
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• 2015

초임계 이산화탄소 발전 시스템 구축을 위해서는 고온, 고압의 열악한 환경에 노출되는 터빈, 열 교환기, 압축기와 같은 핵심 부품들의 내식성 평가는 반드시 수행되어야 한다. 이를 위해 후보소재 3종 Ferritic-Martensitic Steel (T92), Austenitic Steel (SS316L), Ni-based Alloy (IN738LC)를 선정하여 고온의 유사 초임계 $CO_2$ 발전 환경에서의 내식성 평가를 진행하였다. $600^{\circ}C$, $700^{\circ}C$의 2개의 온도 구간에서 $CO_2$ 분위기를 조성하여 800 시간 동안 노출시킨 뒤, Weight Change, Surface Morphology, Cross Section, Composition을 분석하였다. Cr-rich Protective Layer를 형성하는 Ni-based Alloy와 Fe/Cr-rich oxide를 형성하는 Austenitic Steel은 우수한 부식 저항성을 보인 반면에 Ferritic-Martensitic Steel은 높은 Weight Change와 Fe-rich Non-Protective Oxide가 관찰되어 상대적으로 낮은 부식 저항성을 보였다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.12-19
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• 2015

As a measure of improving the mechanical properties of a diffusion bonded joint of a ferritic/martensitic steel (FMS), the post-bonding heat treatment (PBHT) is applied. In the temperature range of normalizing condition ($950-1,050^{\circ}C$), diffusion bonding is employed with compressive stress (6 MPa). Due to the martensite structure distributed in the matrix, Vicker's hardness values of the as-bonded are much higher than those of the as-received. Through the PBHT for 1 h at $720^{\circ}C$, hardness values are recovered to as low as those of the as-received condition. Also, tensile properties of PBHT are similar to those of the as-received at up to the test temperature of $550^{\circ}C$, when the diffusion bonding is carried out over $1,000^{\circ}C$. Based on the creep-rupture testing performed at $650^{\circ}C$ in air environment, the joint efficiency of the PBHTed specimens is about 80% in, which is higher than that of the as-bonded specimens.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1365-1372
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• 2019

The effects of Y and Ti on the microstructure stability and tensile properties of the reduced activation ferritic/martensitic steel have been investigated. The addition of Y and Ti affected the prior austenite grain size due to the pinning of the inclusions. Ti addition of 0.008 wt% to the steel was intended to promote the precipitation of nano-sized carbides with a high resistance to coarsening. 8Ti14Y exhibited a higher yield strength and a lower DBTT than the other alloys due to the fine grain size and additional precipitation hardening by (Ti, Ta)-rich MX. After thermal exposure at $550^{\circ}C$ for 1500 h, yield strength was dropped significantly in exposed 0Ti13Y. On the contrary, a lower reduction of YS was observed in 8Ti14Y. The $M_{23}C_6$ in 0Ti13Y and 8Ti14Y and MX in 25Ti14Y and 39Ti15Y coarsened seriously during ageing, which could be responsible for the reduction of the tensile properties of alloys.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.132-139
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• 2018

Nanocrystalline reduced activation ferritic/martensitic (RAFM) steel samples were prepared using surface mechanical attrition treatment (SMAT). Un-SMATed and SMATed reduced activation ferritic/martensitic samples were irradiated by helium ions at $200^{\circ}C$ and $350^{\circ}C$ with 2 dpa and 8 dpa, respectively, to investigate the effects of grain boundaries (GBs) and temperature on the formation of He bubbles during irradiation. Experimental results show that He bubbles are preferentially trapped at GBs in all the irradiated samples. Bubble denuded zones are clearly observed near the GBs at $350^{\circ}C$, whereas the bubble denuded zones are not obvious in the samples irradiated at $200^{\circ}C$. The average bubble size increases and the bubble density decreases with an increasing irradiation temperature from $200^{\circ}C$ to $350^{\circ}C$. Both the average size and density of the bubbles increase with an increasing irradiation dose from 2 dpa to 8 dpa. Bubbles with smaller size and lower density were observed in the SMATed samples but not in the un-SMATed samples irradiated in the same conditions, which indicate that GBs play an important role during irradiation, and sink strength increases as grain size decreases.

• Korean Journal of Metals and Materials
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• v.48 no.8
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• pp.691-698
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• 2010

A study has been carried out to evaluate the interaction behavior between a lanthanide element and clad material in order to analyze the effect of the lanthanide element on the fuel cladding chemical interaction (FCCI). A diffusion couple test between Misch metal (70Ce-30La) and ferritic-martensitic steel (Gr.92) was performed at $660^{\circ}C$, followed by a microstructural analysis of the coupled sample. The results showed that Ce in the Misch metal, rather than La, reacted with the ferritic-martensitic steel (FMS) to form an interaction layer that penetrated the clad thickness. Fe diffused outside the clad interface to form an $Fe_2Ce$ compound, leaving a depletion of Fe caused by excess diffusion as well as by the formation of Cr-rich precipitation inside the interaction layer. The rate of growth followed the cubic rate law, which indicated that Fe depletion was caused by the diffusion of Fe and that the associated Cr-rich phase formation controlled the whole diffusion process.

• Korean Journal of Metals and Materials
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• v.47 no.10
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• pp.621-628
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• 2009

Thermomechanical processes were carried out to evaluate their effects on the mechanical and the microstructural property of a ferritic-martensitic steel. Modified 9Cr-1Mo steels were hot-rolled at a temperature of either $780^{\circ}C$ or $850^{\circ}C$ after normalizing at $1050^{\circ}C$ and then were air-cooled. Continuous annealing at $850^{\circ}C$ for 2 hours immediately after the hot rolling was also performed and they were compared to the specimens without thermomechanical process. The result showed that there were little differences between the hot rolled specimens in terms of the precipitation density and size. However, V content inside the MX precipitates increased in the case of the specimen rolled at $850^{\circ}C$. The application of the continuous annealing induced coarsening of the Nb-rich MX precipitation as well as an increase in the amount of V-rich MX precipitation, which is expected to enhance high temperature mechanical properties of the ferritic-martensitic steel.

• Korean Journal of Metals and Materials
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• v.47 no.4
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• pp.209-216
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• 2009

It was evaluated that the microstructural and mechanical properties of Ta-bearing 9Cr-0.5Mo-2W ferritic/martensitic experimental steels. All the experimental steels showed the tempered martensitic microstructures, and $M_{23}C_6$ carbides, whose sizes were ranged from 200 to 300 nm, were easily observed at both boundaries of the prior austenite grain and the martensite lath. In addition, a relatively large Nb-rich MX carbonitrides were intermittently detected at the prior austenite grain boundaries, whereas a lot of Vrich MX carbonitrides, whose mean diameter was less than 50 nm, were observed randomly at both boundaries. Ta was mainly incorporated into the V-rich MX carbonitrides rather than the Nb-rich ones and their content was spanned from 5 to 20 at.%. Ta contents within the MX precipitates also increased as the content of Ta increased. Because the Ta addition into the steels would be attributed to the precipitation strengthening, solid solution strengthening and lath width reduction, it was shown that the mechanical properties, including hardness, tensile strength and creep rate of the 9%Cr-0.5Mo-2W steels were improved by the increase of Ta content. Especially, 9Cr-0.5Mo-2W-0.3V-0.05Nb-0.14Ta steel was revealed to be relatively excellent in the application for the SFR fuel cladding.