Effects of pellet microstructure on irradiation behavior of UO2 fuel

Yuda, Ryoichi; Harada, Hideto; Hirai, Masahiro; Hosokawa, Takanari; Une, Katsumi; Kashibe, Shinji; Shimizu, Shunichi; Kubo, Takashi

doi:10.1016/s0022-3115(97)00122-0

Cited by 10 publications

(10 citation statements)

References 15 publications

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“…This is because many of the ceramic alloys are typically used in fuel cladding and must be able to withstand critical core scenarios. Other studies [22][23][24][25][26][27][28][29][30][31] showed that with an increase in neutron flux, the materials showed a temporary increase in hardness as dislocations helped in cases of work-hardening induced by shockwaves and other phenomena. However, the increase in embrittlement resulted in a gradual reduction in yield strength (see Fig.…”

Section: B Ceramic Materialsmentioning

confidence: 98%

A Cohesive Model to Predict Mechanical Responses in Novel Nuclear Fusion Materials and Designs Using a Combined Computational and Empirical Approach

Rodriguez¹,

Cassibry²,

Marlar³

et al. 2014

50th AIAA/ASME/SAE/ASEE Joint Propulsion Conference

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The intent of this paper is to establish the groundwork for additional research into applying nuclear-based materials for space exploration, and attempt to link computational simulations of material exposure to fusion environments with analytical and empirical-based testing of structural and shielding components in a comprehensive, multiphysics scheme. While a special emphasis is placed on the effects of neutron activation and developing a cohesive model that links computational approaches to analytical and empirical tests (especially those used in linear-elastic fracture mechanics and nonlinear approaches), this report will also examine the effects of other sources of damage such as erosion and ablation of electrode structures, which also contribute to reducing component lifetime. Nomenclature A= atomic number α = incident angle B, B 0 = creep compliance b = fatigue strength exponent C 0 = swelling-enhanced creep coefficient c S = solid phase heat capacity D = creep rate ε = material constant derived from the energy needed to remove a unit volume of material ε A = material strain ε' f = fatigue ductility coefficient I = fusion yield scaling coefficient I s = material constant for stress-to-rupture calculations K = particle's vertical velocity component normal to the surface k, k s = thermal conductivity k l , k v = thermal ablation material parameters M = mass of particle N f = number of cycles to failure q = heat flux q * = melting threshold factor φ = material constant relating to the average depth of impact R = Helium to dpa ratio ρ s = density of solid phase S R = Parameter relating rupture stress to time σ = electrical conductivity σ' f = fatigue strength coefficient T mp = melting temperature V = particle velocity V e = volumetric loss W = volume of material lost

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Section: B Ceramic Materialsmentioning

confidence: 98%

A Cohesive Model to Predict Mechanical Responses in Novel Nuclear Fusion Materials and Designs Using a Combined Computational and Empirical Approach

Rodriguez¹,

Cassibry²,

Marlar³

et al. 2014

50th AIAA/ASME/SAE/ASEE Joint Propulsion Conference

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“…Interior bubbles will migrate by surface diffusion, bulk diffusion or vapour transport. Fuel swelling has been measured for normal operating conditions and for transients up to 2270 K in various experiments with fuel swelling measurements of the grain face area, number of pores, bubble density per unit area, fractional coverage, volumetric swelling due to gaseous and solid fission products, bubble coalescence, and fraction of bubbles vented to grain-edge tunnels [86][87][88][89][90][91][92][93][94][95][96]. The main contribution to fuel swelling under CANDU LOCA conditions arises from grainboundary bubbles.…”

Section: Uranium Dioxide Restructuring and Relocationmentioning

confidence: 99%

Overview of high-temperature fuel behaviour with relevance to CANDU fuel

Lewis

Iglesias

Dickson

et al. 2009

Journal of Nuclear Materials

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“…If the volume of the fuel exceeds that of the cladding, or the cladding experiences excessive strain from the swollen fuel, the cladding may rupture, resulting in reactor downtime. The solution to this issue is to improve the UO2 fuel by promoting the growth of larger grains and lengthening the diffusion pathway of fission products to the grain boundaries during fission [2]. This is achieved by doping UO2 with additives; the most extensively applied are Cr2O3 [1,3,4], Al2O3 [5] and a mixture thereof [1,6].…”

Section: Introductionmentioning

confidence: 99%

Hot Isostatic Pressing (HIP): A novel method to prepare Cr-doped UO2 nuclear fuel

et al. 2020

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The addition of Cr2O3 to modern UO2 fuel modifies the microstructure so that, through the generation of larger grains during fission, a higher proportion of fission gases can be accommodated. This reduces the pellet-cladding mechanical interaction of the fuel rods, allowing the fuels to be “burned” for longer than traditional UO2 fuel, thus maximising the energy obtained. We here describe the preparation of UO2 and Cr-doped UO2 using Hot Isostatic Pressing (HIP), as a potential method for fuel fabrication, and for development of analogue materials for spent nuclear fuel research. Characterization of the synthesised materials confirmed that high density UO2 was successfully formed, and that Cr was present as particles at grain boundaries and also within the UO2 matrix, possibly in a reduced form due to the processing conditions. In contrast to studies of Cr-doped UO2 synthesised by other methods, no significant changes to the grain size were observed in the presence of Cr.

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Effects of pellet microstructure on irradiation behavior of UO2 fuel

Cited by 10 publications

References 15 publications

A Cohesive Model to Predict Mechanical Responses in Novel Nuclear Fusion Materials and Designs Using a Combined Computational and Empirical Approach

A Cohesive Model to Predict Mechanical Responses in Novel Nuclear Fusion Materials and Designs Using a Combined Computational and Empirical Approach

Overview of high-temperature fuel behaviour with relevance to CANDU fuel

Hot Isostatic Pressing (HIP): A novel method to prepare Cr-doped UO2 nuclear fuel

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