• 동력기계공학회지
• /
• 제7권2호
• /
• pp.56-60
• /
• 2003

Recently the plastic fuel tube is usually used to reduce production cost and weight in automobiles. These days, material used to plastic fuel tube is the polyamide12. The fuel tube is made of the PA12. Post deformation of the tube has been changed by environmental temperature. So, it is important to prevent post deformation. The experiment is performed to investigate post deformation of the tube produced by each bending process. In this study, the results we obtained are used to bending process system for post deformation as the environmental temperature of the tube. It turned out that the method of steam heating and air cooling was shown less deformation than other methods.

• 동력기계공학회지
• /
• 제7권2호
• /
• pp.51-55
• /
• 2003

Recently the requirements for light weight and high performance of the automobile have increased. Especially, the plastic fuel tube makers have made their efforts to dove]op the various plastic fuel tube module with not only dimensional accuracy but also cost competitiveness. The experiment is performed to investigate spring backs for PA12 plastic fuel tubes in case of compound bending. In the experiment, steam bending process is adopted as bending method. In this study, the results we obtained are used to design the bending fixtures and the compound bending system.

• 한국정밀공학회:학술대회논문집
• /
• 한국정밀공학회 1997년도 추계학술대회 논문집
• /
• pp.825-828
• /
• 1997

The market share of plastic fuel tube in automobile part is now growing rapidly. Especially fuel tube makers have had their efforts to develop tube module not only with dimensional accuracy, spring back and cost competitiveness. In this study, we used steam bending process for heat relaxation on PA12 plastic fuel tube's 128 types experimental bending conditions. we present the results of this process system in term of dimensional accuracy, and spring back.

• Design & Manufacturing
• /
• 제1권1호
• /
• pp.1-5
• /
• 2007

Double injection molding process is very efficient molding-method for molding the products which is consist of multi-materials. Fuel-tube holder which is necessary for automobil power train and circulation systems is composed of plastic and rubber materials to minimize the vibration and pulsation noises. In existing process, fuel-tube holder was made by the insert molding process or assembly process after molding. If fuel-tube holder is manufactured by double injection molding process, it may be realize to improve the product quality, efficiency of molding-process and retrenchment of manufacturing cost. In this study, for manufacturing fuel-tube holder by double injection molding process, the analysis of joining characteristics between PA6(polyamide 6) and TPE(thermoplastic elastomer) was executed and the double injectin mold for molding fuel-tube holder with core toggle mechanism was fabricated. Finally, fuel-tube holder was molding using fabricated double injection mold.

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

• 한국윤활학회:학술대회논문집
• /
• 한국윤활학회 2002년도 제35회 춘계학술대회
• /
• pp.88-95
• /
• 2002

Fretting of fuel rod cladding material, Zircaloy-4 Tube, in PWR nuclear power plants must be reduced and avoided. Nowadays the introduction of surface treatments or coatings is expected to bean ideal solution to fretting damage since fretting is closely related to wear, corrosion and fatigue. Therefore, in this study the fretting wear experiment was peformed using TiAIN coated Zircaloy-4 tube as the fuel rod cladding and uncoated Zircaioy-4 tube as one of grids, especially concentrating on the sliding component. Fretting wear resistance of TiAIN coated Zircaloy-4 tubes was improved compared with that of TiN coated tubes and uncoated tubes and the fretting wear mechanisms were delamination and plastic flow following by brittle fracture at lower slip amplitude but severe oxidation and spallation of oxidative layer at higher slip amplitude.

• 한국윤활학회:학술대회논문집
• /
• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
• /
• pp.185-186
• /
• 2002

Fretting of fuel rod cladding material, Zircaloy-4 tube, in PWR nuclear power plants must be reduced and avoided. Nowadays the introduction of surface treatments or coatings is expected to be an ideal solution to fretting damage since fretting is closely related to wear. corrosion and fatigue. Therefore. in this study the fretting wear experiment was performed using TiAlN coated Zircaloy-4 tube as the fuel rod cladding and uncoated Zircaloy-4 as on of grids, especially concentrating on the sliding component. Fretting wear resistance of TiAlN coated Zircaloy-4 tubes was improved compared with that of TiN coated tubes and uncoated tubes and fretting wear mechanisms were brittle fracture and plastic flow at lower slip amplitude but severe oxidation and spallation of oxidative layer at higher ship amplitude.

• 방사성폐기물학회지
• /
• 제19권3호
• /
• pp.307-322
• /
• 2021

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

• Nuclear Engineering and Technology
• /
• 제29권2호
• /
• pp.175-180
• /
• 1997

It has been known that eater-side corrosion of fuel rods in nuclear reactor is accompanied with the metallic loss of wall thickness and hydrogen pickup in the fuel dadding tube. The fuel dad corrosion is one of the major factors to be controlled to maintain the fuel integrity during reactor operation. An oxide later thickness measuring device equipped with ECT probe system was developed by KAERI, and whose performance test was carried out in NDT(Non-destructive Test) hot-cell or PIE(Post Irradiation Examination) Facility. At first, the calibration/performance test was executed for the unirradiated standard specimen rod fabricated with several kinds of plastic thin films whose thickness ore predetermined, and the result of which showed a good precision within 10% of discrepancy. And then, hot test us peformed for the irradiated fuel rod selectively extracted from J44 fuel assembly discharged from Kori Unit-2. The data obtained with this device were compared with the metallographic result obtained from destructive examination in PIEF hot-cell on the same fuel rod to verify the validity of the measurement data.

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