• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1740-1747
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• 2017

To investigate allowable peak cladding temperature and hoop stress for maintenance of cladding integrity during interim-dry storage and subsequent transport, zirconium alloy cladding tubes were hydrogen-charged to generate 250 ppm and 500 ppm hydrogen contents, simulating spent nuclear fuel degradation. The hydrogen-charged specimens were heated to four peak temperatures of $250^{\circ}C$, $300^{\circ}C$, $350^{\circ}C$, and $400^{\circ}C$, and then cooled to room temperature at cooling rates of $0.3^{\circ}C/min$ under three tensile hoop stresses of 80 MPa, 100 MPa, and 120 MPa. The cool-down specimens showed that high peak heat-up temperature led to lower hydrogen content and that larger tensile hoop stress generated larger radial hydride fraction and consequently lower plastic elongation. Based on these out-of-pile cladding tube test results only, it may be said that peak cladding temperature should be limited to a level < $250^{\circ}C$, regardless of the cladding hoop stress, to ensure cladding integrity during interim-dry storage and subsequent transport.

• 한국소성가공학회:학술대회논문집
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• 한국소성가공학회 2000년도 춘계학술대회논문집
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• pp.171-178
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• 2000

In order to evaluate the tensile behaviors of hoop direction for the nuclear fuel cladding tubes the shapes of specimen and jig fixtures for the ring test are decided with various conditions under the elastic-large plastic deformations. The axial displacement of the jig cylinders is converted to the circumferential direction elongations of specimen. The stress distributions on specimen are depended on the radii and locations of specimen and jig size and central angle. Therefore we calculated the stress distributions and decided the optimum shapes to get the uniform stress in the area of specimen gage length. Form the analysis the stress distributions in gate area are reviewed with the radii and location of specimen notch and the central angle of jig cylinder,. The optimum shapes of specimen and jig are proposed to the clad tube having 10.62 mm in diameter and 0.63mm in thickness for 16x16 PWR nuclear fuel assembly.

• 한국재료학회지
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• 제12권4호
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• pp.259-263
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• 2002

Three different ribbons of amorphous $Til_{1-x}Be_x$ alloys such as $Ti_{0.59}Be_{0.41},\;Ti_{0.61}Be_{0.39}\;and\;Ti_{0.63}Be_{0.37}$ were made by melt-spinning method to be used as brazing filler metals for joining Zircaloy-4 nuclear fuel cladding tubes, and their crystallization behavior as well as microstructure of the brazed zone were examined. The crystallization behavior was investigated in teams of thermal stability, crystallization temperature and activation energy. The crystallization of the $Ti_{1-x}Be_x$ alloys proceeded in two steps by the formation of ${\alpha}$-Ti at a lower temperature and of TiBe at a higher temperature. The crystallization temperature and activation energy of $Ti_{1-x}Be_x$ alloys were higher and larger than those of $Zr_{1-x}Be_x$ alloys and PVD Be. Those resulted thinner joining layer with $Ti_{1-x}Be_x$ alloys, which kept sound thickness of Zircaloy-4 nuclear fuel cladding tubes after brazing. But in the brazed zones made by $Ti_{1-x}Be_x$ filler metals, a little solid-solution layers composed of Zr and Ti were formed toward the Zr cladding tube and Zr was detected in the brazed zones. Microstructure of brazed zone was changed from globular to dentrite with decreasing Be content in the $Ti_{1-x}Be_x$ filler metal.

• 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.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.109-116
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• 2023

This study presents a CRUD modelling to simulate the thermal resistance behavior of CRUD, deposited on the surface of a cladding tube of a fuel assembly. When heat produced from fuels transfers to a coolant through a cladding tube, the CRUD acting as an additional thermal resistance is expressed as two layers, i.e., a solid oxide layer and an imaginary fluid layer, which are added to the experimental tube's heat structure of the MARS-KS input data. The validation calculation for the experiments performed in UNIST-DISNY facility showed that the center and surface temperatures of the cladding tube increased as the porosity and the steam amount inside pores of the CRUD got higher. In addition, the temperature gradient in the imaginary fluid layer was calculated to be larger than that in the solid oxide part, indicating that the steam amount inside the layer acted more largely as thermal resistance. It was also evaluated through sensitivity calculations that the cladding tube temperature was more sensitive to the CRUD porosity and the steam amount in pores than to the inlet flow rate of the coolant.

• Nuclear Engineering and Technology
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• 제40권1호
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• pp.21-36
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• 2008

High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.425-431
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• 2018

Different methods of axial and tangential testing and various sample geometries were investigated, and new test geometries were designed to determine the ultimate tensile strength of zirconium cladding tubes. The finite element method was used to model the tensile tests, and the results of the simulations were evaluated. Axial and tangential tensile tests were performed on as-received and machined fuel cladding tube samples of both E110 and E110G Russian zirconium alloys at room temperature to compare their ultimate tensile strengths and the different sample preparation methods.

• Nuclear Engineering and Technology
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• 제19권1호
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• pp.22-33
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• 1987

원자력 발전소에 있어서 정상가동 상태나 이상동작시에 핵연료 피복관의 건전성 확보와 연관하여 피복재의 항복거동은 중요한 문제이다. 급격한 출력상승 상황에서 이산화 우라늄 소결체와 피복관 사이의 노내 조사거동의 차이는 소결체와 피복관 사이에 Contact Pressure를 야기 시킨다. 만일 이 Contact Pressure가 Zircaloy 피복관의 Yield Pressure에 도달하면 피복관에는 영구변형이 일어난다. 이 변형은 원자로의 출력이 정상상태로 회복되더라도 존재하므로 소결체와 피복관 사이의 Gap을 증대시킨다. 이러한 상황을 묘사하기 위해 본 논문에서는 구리 Mandrel과 Zircaloy사이의 열팽창 차이를 이용하는 Mandrel 팽창 실험을 실행했다. 실험 결과 측정된 Zircaloy 피복관의 외경 팽창치와 본 논문에서 유도된 수학적 관계식들을 이용하여 온도에 따른 Zircaloy 피복관의 내부항복압력과 항복응력, 피복재의 항복에 따른 핵연료 소결체와 피복관 사이의 Gap 증대를 구하고, 항복 거동에 따른 온도의 영향을 보기 위해 항복과정의 활성화 에너지를 구했다. 본 실험과 분석에서 얻어진 이들 결과들은 다른 실험 결과들과 상당히 일치하였으며, 이것으로 볼 때 본 논문에서 유도된 관계식들과 Mandrel 팽창 실험이 Zircaloy 피복관의 항복거동과 Gap Expansion 측정에 신뢰성이 있음을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4066-4076
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• 2023

Pressure resistance welding is usually used to seal the connection between the cladding tube and the end plug made of zirconium alloy. The seal welded joint has a direct effect on the service performance of the fuel rod cladding structure. In this paper, the pressure resistance welded joints of zirconium alloy tube-plug structure were obtained by thermal-mechanical simulation experiments. The microstructure and microhardness of the joints were both analyzed. The effect of processing parameters on the microstructure was studied in detail. The results showed that there was no β-Zr phase observed in the joint, and no obvious element segregation. There were different types of Widmanstätten structure in the thermo-mechanically affected zone (TMAZ) and heat affected zone (HAZ) of the cladding tube and the end plug joint because of the low cooling rate. Some part of the grains in the joint grew up due to overheating. Its size was about 2.8 times that of the base metal grains. Due to the high dislocation density and texture evolution, the microhardnesses of TMAZ and HAZ were both significantly higher than that of the base metal, and the microhardness of the TMAZ was the highest. With the increasing of welding temperature, the proportion of recrystallization in TMAZ decreased, which was caused by the increasing of strain rate and dislocation annihilation.

• Tribology and Lubricants
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• 제25권4호
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• pp.250-255
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• 2009

Oxidation characteristics of Zirlo and Zircaloy-4 tubes, which are widely used as nuclear power fuel cladding, are studied in steam environment up to $1200^{\circ}C$. Oxidation resistances are compared in terms of the mass increase due to the absorption of oxygen. The evolution of microscopic structure accompanied with the oxidation process is investigated. Also, the influence of oxidation on the fretting wear characteristics of the tubes is studied. Piezo-electrically actuated rig is employed to fret the tubes with cross-contacting arrangement. Wear scar is observed and measured, by using microscopes and a 3D-profiler. The results of fretting wear are quantified in terms of scar size, wear volume and wear coefficient, and compared for the three different tube materials of oxidated Zirlo, virgin Zirlo and Zircaloy-4.