• Nuclear Engineering and Technology
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• 제40권2호
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• pp.139-146
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• 2008

This paper presents two of the fuel channel designs being considered for the CANDU-SCWR, a pressure-tube type supercritical water cooled reactor. The first is an insulated pressure tube design. The pressure tube is thermally insulated from the hot coolant by a porous ceramic insulator. Each pressure tube is in direct contact with the moderator, which operates at an average temperature of about $80^{\circ}C$. The low temperature allows zirconium alloys to be used. A perforated metal liner protects the insulator from being damaged by the fuel bundles and erosion by the coolant. The coolant pressure is transmitted through the perforated metal liner and insulator and applied directly to the pressure tube. The second is a re-entrant design. The fuel channel consists of two concentric tubes, and a calandria tube that separates them from the moderator. The coolant enters between the annulus of the two concentric fuel channel tubes, then exits the fuel channel through the inner tube, where the fuel bundles reside. The outer tube bears the coolant pressure and its temperature will be the same as the coolant inlet temperature, ${\sim}350^{\circ}C$. Advantages and disadvantages of these designs and the material requirements are discussed.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2288-2297
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• 2023

Current designs of both large reactor units and small modular reactors utilize a nuclear fuel with increasing enrichment. This increasing demand for better nuclear fuel utilization is a challenge for nuclear fuel handling facilities. The operation with higher enriched fuels leads to reduced reserves to legislative and safety criticality limits of spent fuel transport, storage and final disposal facilities. Design changes in these facilities are restricted due to a boron content in steel and aluminum alloys that are limited by rolling, extrusion, welding and other manufacturing processes. One possible solution for spent fuel pools and casks is the burnup credit method that allows decreasing very high safety margins associated with the fresh fuel assumption in spent fuel facilities. This solution can be supplemented or replaced by an alternative solution based on placing the neutron absorber material directly into the fuel assembly, where its efficiency is higher than between fuel assemblies. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the fuel assemblies. The paper summarizes possibilities of fixed neutron absorbers for various nuclear fuel and fuel handling facilities. Moreover, an absorber material was optimized to propose alternative options to boron. Multiple effective absorbers that do not require steel or aluminum alloy compatibility are discussed because fixed absorbers are placed inside zirconium or steel cladding.

• 방사성폐기물학회지
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• 제17권2호
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• pp.127-137
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• 2019

삼불화알루미늄($AlF_3$)이 포함된 염화물-불화물 혼합 용융염에서 ZIRLO 튜브를 이용한 지르코늄 전해정련공정을 실증하였다. 순환 전압전류실험 결과, $AlF_3$의 농도가 증가함에 따라 금속환원의 개시 전위가 일정하게 증가하고 지르코늄-알루미늄 합금형성과 관련된 추가적인 peak의 크기가 점차 증가하는 것으로 나타났다. 전류조절 전착법과 달리, -1.2 V의 일정전위에서 수행한 지르코늄 전해정련에서 방사형 판 구조의 지르코늄 성장이 염의 상단 표면에서 확연하게 나타났으며, 전착물 지름의 크기는 $AlF_3$의 농도에 따라 점차 증가하는 것으로 나타났다. 주사전자현미경(SEM)과 에너지 분산 X선 분광기(EDX)와 X선 광전자 분광기(XPS)를 이용하여 판 구조의 지르코늄 전착물을 분석한 결과, 극미량의 알루미늄이 지르코늄-알루미늄 합금 형태로 존재하며, 전착물의 상단과 하단 간에 서로 다른 화학성분구조를 갖는 것으로 나타났다. $AlF_3$의 첨가는 전착물 내 잔류염 양을 줄이고, 지르코늄 회수를 위한 전류효율을 향상시키는 데 효과적인 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제46권2호
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• pp.169-182
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• 2014

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.16-25
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• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.416-424
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• 2018

Microstructure of oxide formed on Zr-Nb-Sn tube sample was intensively examined by scanning transmission electron microscopy after exposure to simulated primary water chemistry conditions of various concentrations of Zn (0 or 30 ppb) and dissolved hydrogen ($H_2$) (30 or 50 cc/kg) for various durations without applying desirable heat flux. Microstructural analysis indicated that there was no noticeable change in the microstructure of the oxide corresponding to water chemistry changes within the test duration of 100 days (pretransition stage) and no significant difference in the overall thickness of the oxide layer. Equiaxed grains with nano-size pores along the grain boundaries and microcracks were dominant near the water/oxide interface, regardless of water chemistry conditions. As the metal/oxide interface was approached, the number of pores tended to decrease. However, there was no significant effect of $H_2$ concentration between 30 cc/kg and 50 cc/kg on the corrosion of the oxide after free immersion in water at $360^{\circ}C$. The adsorption of Zn on the cladding surface was observed by X-ray photoelectron spectroscopy and detected as ZnO on the outer oxide surface. From the perspective of $OH^-$ ion diffusion and porosity formation, the absence of noticeable effects was discussed further.

• Nuclear Engineering and Technology
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• 제8권1호
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• pp.19-27
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• 1976

질칼로이-4의 가공시효(strain ageing) 현상을 173-575$^{\circ}C$의 온도구간에서 소입된 시편과 소둔된 시편에 대해 각각 조사하였다. 소입된 질칼로이-4의 가공시효가 175-50$0^{\circ}C$구간체서 난타났으며, 3$25^{\circ}C$에서 극대값의 가공시효가 조사였었다. 반면 소둔된 시편의 경우엔 175-575$^{\circ}C$ 구간에 걸쳐 가공시효가 나타났으며, 이 온도구간 사이에 3$25^{\circ}C$ 및 45$0^{\circ}C$의 두 지점에서 극대값이 나타나는 것을 확인했다. 소입된 경우와 소둔된 경우에 모두 3$25^{\circ}C$에서 보인 가공시효의 극대값은 시효동안에 전위조직 속으로 침입형 산소원자들의 응집에 의한 결과로 고려되며 45$0^{\circ}C$에서 보이는 극대값은 철원자와 전위간의 작용에 기인된 것으로 해석된다. 질코늄 합금의 30$0^{\circ}C$ 근처에서 보이는 가공시효값은 첨가된 산소량의 평방근에 비례한다는 것이 확인됐다.

• 한국표면공학회:학술대회논문집
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• 한국표면공학회 2000년도 춘계학술발표회 초록집
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• pp.1-1
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• 2000

The results of the modification of metal materials treated by high temperature pulst:d plasma fluxes (HTlPPF) with a specific power of incident flux changing in the $(3...100)10^5{]\;}W/cm^2$ range and a pulse duration lying from 15 to $50{\;}\mu\textrm{s}$ have been presented. The results of HTPPF action were studied on the stainless steels of 18Cr-l0Ni, 16Cr- 15Ni, 13Cr-2Mo types; on the structural carbon steels of (13...35)Cr, St. 3, St. 20, St. 45 types; on the tool steels of U8, 65G, ShHI5 types, and others; on nickel and high nickel alloy of 20Cr-45Ni type; on zirconium- and vanadium-base alloys and other materials. The microstructure and properties (mechanical, tribological, erosion, and other properties) of modified materials and surface alloying of metals exposed to HTPPF action have been investigated. It was found that the modification of materials by HTPPF resulted in a simultaneous increase of several properties of the treated articles: microhardness of the surface and layers of 40...60 $\mu\textrm{m}$ in depth, tribological characteristics (friction coefficient, wear resistance), mechanical properties ({\sigma_y}, {\;}{\sigma_{0.2}}.{\;}{\sigma_r}) on retention of the initial plasticity ($\delta$), corrosion resistance, radistanation erosion under ion irradiation, and others. The determining factor of the changes observed is the structural-phase modification of the near-surface layers, in particular, the formation of the fine cellular structure in the near-surface layers at a depth of $20{\;}{\mu\textrm{m}}$ with dimension of cells changing in the range from 0.1 to $1., 5{\;}\mu\textrm{m}$, depending on the kind of material, its preliminary treatment, and the parameters of plasma fluxes. The remits obtained have shown the possibility of purposeful surface alloying of metals exposed to HTPPF action over a depth up to 20...45 $\mu\textrm{m}$ and the concentration of alloying element (Ni, Cr, V) up to 20 wt.%. Possible industrial brunches for using the treatment have been also considered, as well as some results on modifying the serial industrial articles by HTPPF.

• 한국압력기기공학회 논문집
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• 제20권1호
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• pp.39-48
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• 2024

Research reactors are operated under ambient temperature and atmospheric pressure, which is much less severe conditions compared to those in typical nuclear power plants. Due to the high temperature, heat resistant materials such as austenite stainless steel should be used for the reactors in typical nuclear power plants. Whereas, as the effect of temperature is low for research reactors, materials with high resistance to neutron irradiation, such as zircaloy and beryllium, are used. Therefore, these conditions should be considered when performing integrity assessment for research reactors. In this study, a computational technique through finite element (FE) analysis was developed considering the operating conditions and materials of research reactor when conducting integrity assessment. Neutron irradiation analysis techniques using thermal expansion analysis were proposed to consider neutron irradiation growth and swelling in zirconium alloys and beryllium. A user subroutine program that can calculate the strain rate induced by neutron irradiation creep was developed for use in the commercial analysis program Abaqus. To validate the proposed technique and the user subroutine, FE analysis results were compared with hand-calculation results, and showed good agreement. Consequently, developed technique and user subroutine are suitable for evaluating structural integrity of research reactors.