• Tribology and Lubricants
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• 제22권3호
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• pp.144-148
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• 2006

The experimental investigation was performed to find the associated changes in characteristics of fretting wear with various water temperatures. The fretting wear tests were carried out using the zirconium alloy tubes and the grids with increasing the water temperature. The tube materials in water of $20^{\circ}C,\;50^{\circ}C\;and\;80^{\circ}C$ were tested with the applied load of 20 N and the relative amplitude of $200{\mu}m$. The worn surfaces were observed by SEM, EDX analysis and 2D surface profiler. As the water temperature increased, the wear volume was decreased, but oxide layer was increased on the worn surface. The abrasive wear mechanism was observed at water temperature of $20^{\circ}C$ and adhesive wear mechanism occurred at water temperature of $50^{\circ}C,\;80^{\circ}C$. As the water temperature increased, surface micro-hardness was decreased, but wear depth and wear width were decreased due to increasing stick phenomenon. Stick regime occurred due to the formation of oxide layer on the worn surface with increasing water temperatures

• 한국재료학회지
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• 제14권10호
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• pp.713-718
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• 2004

To study the dynamic strain aging behavior of Zr-0.4Sn-1.5Nb-0.2Fe sample tube for nuclear fuel cladding in the range of pressurized water reactor (PWR) operation temperature, the tensile tests of the tube specimens, which had been finally heat-treated at $470^{\circ}C\;and\;510^{\circ}C$, had been carried out with the strain rate $1.67{\times}10^{-2}/s\;and\;8.33{\times}10^{-5}/s$ at the various temperatures from room temperature to $500^{\circ}C$. It was observed that the elongation of the specimens got shortened as the temperature increased from $200^{\circ}C\;to\;340^{\circ}C$. The specimens that were finally heat-treated at $470^{\circ}C$ showed a plateau more remarkably on the plot of yield strength-temperature than those heat-treated at $510^{\circ}C$. In the range of $310\sim400^{\circ}C$, the strain rate sensitivity of the specimens finally heat-treated at $510^{\circ}C$ was $30.4\%\sim33.7\%$ lower but the work hardening exponent index of the specimens was a little higher than that without dynamic strain aging effect.

• 열처리공학회지
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• 제34권2호
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• pp.66-74
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• 2021

Cladding material, which can selectively obtain excellent properties of different metals, is a composite material that combines two or more types of dissimilar metals into one plate. The titanium-copper cladding material between titanium which has excellent corrosion resistance and copper which has high thermal and electrical conductivity, are highly valuable composite materials. It can be used as heat exchangers with high conductivity under severe corrosion conditions. In order to apply the clad plate to the heat exchanger, it must be manufactured in the form of a tube and additional welding is required. It is important to select the cladding material manufacturing process and the welding process. The process of manufacturing the cladding material includes extrusion, rolling, and explosive bonding. Among them, the explosive bonding process is suitable for additional welding because no heat-affected zone is formed. In this study TIG welding of the explosive-bonded dissimilar clad plates was successfully performed by butt welding. The microstructures and bonding interface of the welded part were observed, and the effect of the bonding layer at the welding interface and the intermetallic compounds on the mechanical properties and tensile plastic deformation behaviors were analyzed. And also the integrity of TIG-welded dissimilar part was evaluated.

• 비파괴검사학회지
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• 제21권6호
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• pp.650-657
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• 2001

본 논문에서는 핵연료 피복재 튜브를 검사하기 위한 차폐된 관통형 원격장와전류 탐촉자의 설계과정을 설명하고, 이 탐촉자에 의한 결함신호의 특성을 조사하였다. 먼저, 자기 에너지가 튜브 내부로 관통될 수 있도록 여자코일 외부를 전기적으로 절연된 얇은 철 박판을 적층시켜 차폐시켰다. 그리고 유한요소 해석을 통하여 차폐의 효과와 탐상 주파수를 연구하였으며, 센서코일의 위치를 결정하였다. 그러나 이렇게 설계된 탐촉자를 사용하여 예측된 결함신호는 센서코일이 결함을 지날 때의 결함지시가 명확하지 않았으며, 여자코일이 결함을 지날 때의 결함지시도 차폐체로부터의 영향이 나타나는 등 여자코일로부터 자속이 직접적으로 센서코일에 영향을 미친다는 사실을 알게 되었다. 따라서 센서코일도 여자코일과 같은 형태로 차폐시켰는데 이 차폐의 효과는 놀라울 정도로 결함신호의 특성을 향상시켰다. 최종적으로 설계된 탐촉자를 사용하여 수치 모델링을 수행한 결과는 관내삽입 원격장와전류 탐촉자를 사용하였을 때의 신호와 매우 흡사한 신호특성을 보였다. 즉, 위상신호는 내부결함과 외부결함에 대하여 거의 동일한 민감도를 보였으며, 위상신호의 세기와 결함의 깊이 사이에 선형적인 관계가 있음이 관찰되었다.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2001년도 제34회 추계학술대회 개최
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• pp.19-26
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• 2001

The wear characteristic of Zircaloy-4 tube, which is used for a cladding of light water reactor fuel rod, is investigated experimentally. The experiment is conducted with contacting the crossed tube specimens in air as well as in water at room temperature with various combination of contact normal force and sliding distance of reciprocating motion. The contour and the volume of each wear are examined to study the effect of contact condition and environment on wear. As a result, it is found that the wear volume in the water environment is larger than that in the air for all the contact (i.e., force and sliding distance) conditions. However, the wear depth is greater in air than in water if the contact normal force and the sliding distance are larger. These are explained by the ease of detachment of wear particles from the contact surface. On the other hand, workrate model is applied with the contact shear force range measured by our wear tester. Investigated is the correlation between the workrate and the wear volume increase rate of the present experiment. The parabolic curve is found to fit well for the present wear data.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.322-323
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• 2005

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
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• pp.195-196
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• 2002

Studied is fretting wear behaviour of transversely vibrating tube which is supported by springs and dimples. This simulates the fuel rod fretting due to flow-induced vibration in a nuclear reactor. The contact between spacer grid springs and fuel cladding tubes arc brought into focus in this paper. From the mechanical viewpoint, a concave contact shape of spring is considered to perform a wider distribution of the contact stress. Sliding/impacting experiments are conducted in air at room temperature with the conditions of positive contact force and gap existence to accommodate the mechanical condition between the fuel rod and the grid spring during reactor operation. It is found that wear region is separated and wear volume becomes larger as the supporting condition becomes poorer. Spring and dimple cause similar wear.

• 한국안전학회지
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• 제37권4호
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.149-158
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• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• 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