• Nuclear Engineering and Technology
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• 제47권3호
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• pp.253-259
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• 2015

The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ~265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.164-168
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• 2016

Lean duplex stainless steels with 0.087 wt.% gadolinium (Gd) were inert arc-melted and cast in molds of size $10mm{\times}10mm{\times}20mm$. The micro-hardnesses of the rolling direction (RD), transverse direction (TD) and short transverse (ST) direction were $258.5H_V$, $292.3H_V$, and $314.7H_V$, respectively. A 33% cold rolled specimen had the crystallographic texture that (100) pole was mainly concentrated to the normal direction (ND) and (110) pole was concentrated in the center of ND and RD. The corrosion potential and corrosion rate in artificial seawater and $0.1M\;H_2SO_4$ solution were in the range of $105.6-221.6mV_{SHE}$, $0.59-1.06mA/cm^2$, and $4.75-8.25mV_{SHE}$, $0.69-1.68mA/cm^2$, respectively. The friction coefficient and wear loss of the 0.087 w/o Gd-lean duplex stainless steels in artificial seawater were about 67% and 65% lower than in air, whereas the wear efficiency was 22% higher. The corrosion and wear behaviors of the 0.087 w/o Gd-lean duplex stainless steels significantly depended on the Gd phases.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.207-213
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• 2019

Corrosion and mechanical behavior of the hot-rolled 0.5%Gd-0.8%B-stainless steel to develop a spent nuclear fuel storage material was studied in a simulated nuclear waste treatment condition with rolling condition. The austenite and ferrite phases of the 0.5%Gd-0.8%B-stainless steels are about 88:12. The average austenite and ferrite grain size of the plane normal to rolling, transverse and normal directions of the hot rolled specimens are about 5.08, 8.94, 19.35, 23.29, 26.00 and 18.11 [${\mu}m$], respectively. The average micro-hardness of the as-cast specimen is 200.4 Hv, whereas, that of the hot-rolled specimen are 220.1, 204.7 and 203.5 [$H_v$] for the plane normal to RD, TD and ND, respectively. The UTS, YS and elongation of the as-cast and the hot-rolled specimen are 699, 484 [MPa], 34.0%, and 654, 432 [MPa] and 33.3%, respectively. The passivity was observed both for the as-cast and the hot rolled specimens in a simulated nuclear waste solution. The corrosion potential and corrosion rate of the as-casted specimens are $-343[mV_{SHE}]$ and $3.26{\times}10^{-7}[A/cm^2]$, whereas, those of the hot rolled specimens with normal to ND, RD and TD are -630, -512 and -620 [$mV_{SHE}$] and $6.12{\times}10^{-7}$, $1.04{\times}10^{-6}$ and $6.92{\times}10^{-7}[A/cm^2]$, respectively. Corrosion tends to occur preferentially Cr and B rich area.

• 한국표면공학회:학술대회논문집
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• 한국표면공학회 2014년도 추계학술대회 논문집
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• pp.212-212
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• 2014

0.04% Gd-duplex stainless steels (Gd-DSTSs) for neutron absorbing materials were inert arc-melted and poured into a Y-shape block with the size of $100{\times}100{\times}20[mm]$. The Gd-DSTS was hot rolled at $1200^{\circ}C$ followed by cold rolling to have 33% reduction. The average grain sizes of the rolling (RD), transverse (TD) and short transverse (ST) directions were 6, 7, $11{\mu}m$, respectively. The micro-hardnesses of the RD, TD and ST directions were 258.5, 292.3, 314.7 $H_V$, respectively. Corrosion potential and corrosion rate of the cold rolled Gd-duplex stainless steel in aerated artificial sea water and 0.1M $H_2SO_4$ solution were $0.2216V_{SHE}$, $0.0106A/cm^2$, $-0.0825V_{SHE}$, $0.0168A/cm^2$ for RD, $0.2210V_{SHE}$, $0.0077A/cm^2$, $0.0817V_{SHE}$, $0.0092A/cm^2$ for TD, $0.1056V_{SHE}$, $0.0059A/cm^2$, $0.0475V_{SHE}$, $0.0069A/cm^2$ for ST, respectively. The corrosion behavior depended on the texture, which were due to mainly grain boundary and minorly crystallographic texture. Friction coefficient and wear resistance were 2.07 and 0.48 mm, respectively.

• 자원리싸이클링
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• 제21권5호
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• pp.18-30
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• 2012

Zr은 고온에서의 높은 치수안정성, 내식성은 물론 낮은 중성자 흡수단면적을 지녀 원자력산업용 소재 중 1차 방사능 차폐재인 핵연료 피복관으로 사용되며 현재까지 다른 소재로 대체 불가능하다. 하지만 Hf을 정제한 Zr sponge 제조기술은 미국, 프랑스, 러시아만 가지고 있어 원자력의존도가 높은 한국에서는 국가전략물자로 분류 철저히 관리되고 있다. 국내 유통되는 Zr의 대부분은 원자력산업에 사용되어 지며 유통구조는 정제된 Zr합금을 국외로부터 수입하여 tube로 가공 후 핵연료집합체로 제조되고, 그 외 소량이 합금첨가원소 및 폭약재 등 고부가가치 일반산업에 사용된다. 본 논문에서는 Zr 제조기술에 대한 현재산업현황 및 정련기술을 살펴보고, 최근 연구되고 있는 Electrolytic reduction process와 Molten oxide electrolysis와 같은 신 제련기술에 대한 소개 및 Zr recycling의 전반적인 기술소개도 포함하였다.