• 방사성폐기물학회지
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• 제19권3호
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• pp.307-322
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• 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.

• 열처리공학회지
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• 제17권3호
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• pp.151-164
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• 2004

The changes of microstructural and mechanical properties were evaluated for the high burn-up fuel claddings after the neutron irradiation of $1.8{\sim}3.1{\times}10^{20}n/cm^2$ (E>1.0 MEV) in HANARO research reactor. After the irradiation, the spot-type dislocations (a-type dislocations) were easily observed in most claddings, and the density of the dislocations was different depending on the grains and was higher at grain boundaries than within grains. As the final annealing temperature increased, the density of spot-type dislocations increased and the line-type dislocations (c-type dislocations) which was perpendicular to the <0002> direction, appeared sporadically in some claddings. However, the types of precipitates in the fuel claddings after the irradiation were not changed from that in unirradiated claddings. The mechanical properties including the hardness, strength and elongation after the irradiation were changed due to the formation of spot-type dislocations. That is, the increase in hardness and strength as well as the decrease in elongation after the irradiation was occurred simultaneously with increasing the final annealing temperature. Owing to the Nb contribution to the formation of spot-type dislocation during the irradiation, the increase in hardness and strength in higher Nb-contained Zr alloys after the irradiation was higher than that in lower Nb-contained Zr alloys.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.306-312
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• 2018

There is an increasing need in the United States and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations or interim storage sites. Under normal conditions, the Nuclear Regulatory Commission limits cladding temperature to $400^{\circ}C$ for high-burnup (>45 GWd/mtU) fuel, with higher temperatures allowed for low-burnup fuel. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at $400^{\circ}C$. Results were representative of the majority of US light water reactor fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.327-362
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• 2005

The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2047-2053
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• 2020

Single-tube burst tests on hydrogenated Zircaloy-4 nuclear fuel cladding under simulated loss-of-coolant accident are conducted to evaluate the impact of hydrogen on burst parameters. The heating rate and initial pressure are varied from 5 K/s to 150 K/s and 5 bar-80 bar, respectively. The hydrogen concentration in the cladding is in the range of 0-2000 wppm. Burst stress is lower for hydrogenated cladding in α-phase. A significant loss of ductility is observed in α-phase and lower α + β-phase for hydrogenated cladding. However, the burst strain is higher for hydrogenated cladding in β-phase. There is a sigmoidal dependency of rupture area with initial stress and rupture area is larger for hydrogenated cladding. A novel burst stress correlation for hydrogenated Zircaloy-4 cladding has been proposed.

• 방사성폐기물학회지
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• 제21권1호
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• pp.115-147
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• 2023

The Fukushima accident in 2011 revealed some major flaws in traditional nuclear fuel materials under accidental conditions. Thus, the focus of research has shifted toward "accident tolerant fuel" (ATF). The aim of this approach is to develop fuel material solutions that lead to improved reactor safety. The application of protective coatings on the surface of nuclear fuel cladding has been proposed as a near-term solution within the ATF framework. Many coating materials are being developed and evaluated. In this article, an overview of different zirconium-based alloys currently in use in the nuclear industry is provided, and their performances in normal and accidental conditions are discussed. Coating materials proposed by different institutions and organizations, their performances under different conditions simulating nuclear reactor environments are reviewed. The strengths and weaknesses of these coatings are highlighted, and the challenges addressed by different studies are summarized, providing a basis for future research. Finally, technologies and methods used to synthesize thin-film coatings are outlined.

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

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

• 대한기계학회논문집A
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• 제40권5호
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• pp.437-447
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• 2016

원자력 발전소의 반응로에 핵연료 봉으로 이루어진 집합체가 있으며 핵 연료의 연소를 통한 열을 이용하여 발전을 한다. 핵연료 봉은 핵연료와 그를 감싸는 피복관으로 이루어졌으며 연소되는 동안 서로의 상호작용에 대한 해석은 안전성을 평가함에 있어 중요한 사실이다. 본 논문에서는 핵연료와 피복관의 연소 상태에서 기계적 상호작용에 대한 해석 방법에 대하여 제시한다. 온도 해석에 있어서 핵연료와 간극 사이에서의 열전도도가 중요하며 간극 거리와 접촉여부에 따른 접촉 압력이 또한 중요 요소이다. 이에 간극 열전도도는 비결정론적이기 때문에 이를 해결할 수 있는 방법에 대하여 제시했다. 핵 연료의 열팽창에 따른 피복관과의 접촉을 해결하기 위한 계산을 수행하였고 그에 따라 접촉 시 발생하는 응력이 항복함수를 넘어 소성 변형이 일어날 경우 또한 고려하였다. 핵연료의 열팽창에 따라 피복관과 접촉에 의한 소성 변형을 해석하므로 핵연료 봉의 안정성을 평가할 수 있다. 이를 적용하기 위해 3차원 유한요소 모듈을 FORTRAN90을 이용하여 개발하였다. 핵연료와 피복관의 접촉에 의한 탄소성 변형을 주로 다루며 두꺼운 실린더를 통한 간단한 이론 모델을 제시하여 코드에 대해 검증을 실시하였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.58-59
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• 2018