• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• 비파괴검사학회지
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• 제30권4호
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• pp.324-329
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• 2010

원자력발전소에서 사용되고 있는 연료봉은 지르코늄 합금 튜브에 동봉되어 있는 이산화우라늄 펠릿으로 구성되어 있다. 펠릿 표면은 원자로를 가동시키는 동안 국부 핫스팟을 예방하기 위해 튜브로 장전된 후 작은 구멍, 균열, 칩핑 결함이 없어야 한다. 본 논문은 X-선 단층촬영 시뮬레이션을 통하여 핵 연료봉 펠릿의 표면 결함을 검출하기 위한 타당성을 조사하였다. 병렬과 팬빔 여과후 역투영 방법을 이용하여 재구성된 영상은 시뮬레이션 데이터와 MPS(missing pellet surface) 영상데이터의 접근성을 확인하였다.

• 한국재료학회지
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• 제14권6호
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• pp.394-398
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• 2004

A Ti47Al1.7W-3.7Zr alloy was oxidized between $900^{\circ}C$ and $1050^{\circ}C$, and the oxide scales formed were studied. The oxide scales consisted primarily of an outer$TiO_2$ layer, an intermediate $Al_2$$O_3$-rich layer, and an inner mixed ($TiO _2$ + $Al_2$$O_3$) layer. Besides $TiO_2$ and $Al_2$$O_3$, oxidation led to the formation of some $Ti_2$AlN and TiN. Both W and Zr were preferentially segregated below the intermediate $Al_2$$O_3$-rich layer. Tungsten in the oxide scale was present as $WO_3$ and ${Ti}_{x}$$W_{1-x}$, whereas zirconium as monoclic-$ZrO_2$ and tetragonal-$ZrO_2$.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.521-528
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• 2014

The surface modification of engineering materials by laser beam scanning (LBS) allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS) structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 mm, and $Y_2O_3$ particles of $10{\mu}m$ were selected for ODS treatment using LBS. Through the LBS method, the $Y_2O_3$ particles were dispersed in the Zircaloy-4 sheet surface at a thickness of 0.4 mm, which was about 20% when compared to the initial sheet thickness. The mean size of the dispersive particles was 20 nm, and the yield strength of the ODS treated plate at $500^{\circ}C$ was increased more than 65 % when compared to the initial state. This strength increase was caused by dispersive $Y_2O_3$ particles in the matrix and the martensite transformation of Zircaloy-4 matrix by the LBS.

• 한국주조공학회지
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• 제19권5호
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• pp.403-409
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• 1999

Mg-Zn-Zr ternary alloys containing 6wt% Zn and (0, 0.4, 0.6)wt% Zr, which is added for grain refinement, can be cast into complex shape by squeeze casting. The influence of Zn and Zr as additional elements on microstructure and mechanical characteristics is investigated by OM, SEM, WDX, XRD and microvickers hardness measurement. The microstructure of Mg-Zn-Zr alloys consists of primary ${\alpha}-Mg$ and MgZn eutectic compound between dendrites. The grain size is decreased from $136{\mu}m$ to $97\;{\mu}m$ by Zr addition, resulting in that the hardness is increased from 42Hv to 59Hv. Furthermore, the grain size is changed to $83{\beta}$ and the hardness is increased to 65Hv by additional infiltration pressure. These results indicate that the Zr addition and additional infiltration pressure are effective for grain refinement acting as an important factor to increase the hardness. The increment in hardness by the Zr addition is slightly larger than that by the additional infiltration pressure.

• 한국재료학회지
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• 제27권12호
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• pp.672-678
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• 2017

In the production of zirconium cladding tube, a pickling acid solution is used to remove surface contaminants, which generates tons of pickling acid waste. The waste pickling solution is a valuable resource of Hf-free Zr. Many studies have investigated separating the Hf-free Zr source from the waste pickling acid. The results showed that $Ba_2ZrF_8$ precipitates prepared from the waste pickling acid were useful as an electrolyte for the electrorefining of Zr in molten salt. In the present work, electrorefining was performed in a $Ba_2ZrF_8-LiF$ binary electrolyte to recover Zr from a Hf-free CuZr ingot anode prepared by electroreduction. Before electrorefining, two pretreatments are performed. First, electrolyte melting was carried out to determine the eutectic temperature, and second, the electrolyte was treated to eliminate impurities, mainly hydride. After electrorefining, the cathode deposits were analyzed by $O_2$ gas analyzer and SEM-EDX to explore the possibility of recovering nuclear-grade Zr metal. Moreover, the anode was analyzed by SEM-EDX to determine the Zr dissolution depth.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2475-2490
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• 2022

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO₂-Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3164-3182
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• 2023

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

• 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.

• 전기화학회지
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• 제12권1호
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• pp.26-33
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• 2009

The phase-shift method and correlation constants, i.e., the unique electrochemical impedance spectroscopy (EIS) techniques for studying the linear relationship between the behavior ($-{\varphi}$ vs. E) of the phase shift ($90^{\circ}{\geq}-{\varphi}{\geq}0^{\circ}$) for the optimum intermediate frequency and that ($\theta$ vs. E) of the fractional surface coverage ($0{\leq}{\theta}{\leq}1$), have been proposed and verified to determine the Langmuir, Frumkin, and Temkin adsorption isotherms of H and related electrode kinetic and thermodynamic parameters at noble metal (alloy)/aqueous solution interfaces. At a Zr/0.2 M ${H_2}{SO_4}$ aqueous solution interface, the Frumkin and Temkin adsorption isotherms ($\theta$ vs. E), equilibrium constants (K = $1.401{\times}10^{-17}\exp(-3.5{\theta})mol^{-1}$ for the Frumkin and K = $1.401{\times}10^{-16}\exp(8.1{\theta})mol^{-1}$ for the Temkin adsorption isotherm), interaction parameters (g = 3.5 for the Frumkin and g = 8.1 for the Temkin adsorption isotherm), rates of change of the standard free energy (r = $8.7\;kJ\;mol^{-1}$ for g = 3.5 and r = $20\;kJ\;mol^{-1}$ for g = 8.1) of H with $\theta$, and standard free energies ($96.13{\leq}{\Delta}G^0_{\theta}{\leq}104.8\;kJ\;mol^{-1}$ for K = $1.401{\times}10^{-17}\exp(-3.5{\theta})mol^{-1}$ and $0{\leq}{\theta}{\leq}1$ and ($94.44<{\Delta}G^0_{\theta}<106.5\;kJ\;mol^{-1}$ for K = $1.401{\times}10^{-16}\exp(-8.1{\theta})mol^{-1}$ and $0.2<{\theta}<0.8$) of H are determined using the phase-shift method and correlation constants. At 0.2 < $\theta$ < 0.8, the Temkin adsorption isotherm correlating with the Frumkin adsorption isotherm, and vice versa, is readily determined using the correlation constants. The phase-shift method and correlation constants are probably the most accurate, useful, and effective ways to determine the adsorption isotherms of H and related electrode kinetic and thermodynamic parameters at highly corrosion-resistant metal/aqueous solution interfaces.