scholarly journals Properties of the high burnup structure in nuclear light water reactor fuel

2017 ◽  
Vol 105 (11) ◽  
Author(s):  
Thierry Wiss ◽  
Vincenzo V. Rondinella ◽  
Rudy J. M. Konings ◽  
Dragos Staicu ◽  
Dimitrios Papaioannou ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Nuclear Fuel ◽  
Experimental Studies ◽  
Fuel Pellet ◽  
Extended Defects ◽  
Safety Margins ◽  
Fission Gas ◽  
High Burnup ◽  
Formed Structure

AbstractThe formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2–3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings from extensive studies performed at JRC-Karlsruhe during the last 25 years to determine properties and behaviour of the HBS. In particular, microstructural features, thermal transport, fission gas behaviour, and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental studies is that the HBS does not compromise the safety of nuclear fuel during normal operations.

FRACTAL MODEL OF FISSION PRODUCT RELEASE IN NUCLEAR FUEL

2012 ◽  
Vol 23 (09) ◽  
pp. 1250057
Author(s):  
GEDIMINAS STANKUNAS
Keyword(s):  
Diffusion Process ◽  
Nuclear Fuel ◽  
Functional Dependence ◽  
Experimental Studies ◽  
Fractional Diffusion ◽  
Fractal Model ◽  
Fuel Pellet ◽  
Operation Conditions ◽  
Fuel Burn ◽  
Fission Gas

A model of fission gas migration in nuclear fuel pellet is proposed. Diffusion process of fission gas in granular structure of nuclear fuel with presence of inter-granular bubbles in the fuel matrix is simulated by fractional diffusion model. The Grunwald–Letnikov derivative parameter characterizes the influence of porous fuel matrix on the diffusion process of fission gas. A finite-difference method for solving fractional diffusion equations is considered. Numerical solution of diffusion equation shows correlation of fission gas release and Grunwald–Letnikov derivative parameter. Calculated profile of fission gas concentration distribution is similar to that obtained in the experimental studies. Diffusion of fission gas is modeled for real RBMK-1500 fuel operation conditions. A functional dependence of Grunwald–Letnikov derivative parameter with fuel burn-up is established.

scholarly journals Iodine release from high-burnup fuel structures: Separate-effect tests and simulated fuel pellets for better understanding of iodine behaviour in nuclear fuels

MRS Advances ◽  
2021 ◽  
Author(s):  
Janne Heikinheimo ◽  
Teemu Kärkelä ◽  
Václav Tyrpekl ◽  
Matĕj̆ Niz̆n̆anský ◽  
Mélany Gouëllo ◽  
...  
Keyword(s):  
Fission Product ◽  
Nuclear Fuel ◽  
Chemical Species ◽  
Outer Edge ◽  
Fuel Pellet ◽  
Fuel Pellets ◽  
Simulated Fuel ◽  
Fission Gas ◽  
High Burnup ◽  
Iodine Release

Abstract Iodine release modelling of nuclear fuel pellets has major uncertainties that restrict applications in current fuel performance codes. The uncertainties origin from both the chemical behaviour of iodine in the fuel pellet and the release of different chemical species. The structure of nuclear fuel pellet evolves due to neutron and fission product irradiation, thermo-mechanical loads and fission product chemical interactions. This causes extra challenges for the fuel behaviour modelling. After sufficient amount of irradiation, a new type of structure starts forming at the cylindrical pellet outer edge. The porous structure is called high-burnup structure or rim structure. The effects of high-burnup structure on fuel behaviour become more pronounced with increasing burnup. As the phenomena in the nuclear fuel pellet are diverse, experiments with simulated fuel pellets can help in understanding and limiting the problem at hand. As fission gas or iodine release behaviour from high-burnup structure is not fully understood, the current preliminary study focuses on (i) sintering of porous fuel samples with Cs and I, (ii) measurements of released species during the annealing experiments and (iii) interpretation of the iodine release results with the scope of current fission gas release models. Graphical abstract

Characterization of Mechanical Properties of Nuclear Waste Glasses

1981 ◽  
Vol 11 ◽  
Author(s):  
Herbert Richter ◽  
Peter Offermann
Keyword(s):  
Mechanical Properties ◽  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Waste Disposal ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
Safety Barrier

Part of nuclear fuel cycle waste is highly dangerous, and must be safely isolated from people. Although the site of the final waste disposal must be the main safety barrier, the form of the waste and its properties are also important considerations.

Detailed Characterization of LWR Fuel Rods for the U.S. Civilian Radioactive Waste Management Program

1988 ◽  
Vol 127 ◽  
Author(s):  
Richard J. Guenther ◽  
D. E. Blahnik ◽  
T. K. Campbell ◽  
U. P. Jenquin ◽  
J. E. Mendel ◽  
...  
Keyword(s):  
Oxide Thickness ◽  
Management Program ◽  
Light Water Reactor ◽  
Spent Fuel ◽  
Detailed Characterization ◽  
High Release ◽  
Fission Gas ◽  
High Burnup ◽  
Scan Data

ABSTRACTCharacterizations are being conducted on light-water reactor (LWR) spent fuel with peak burnups from ∼31 to 46 MWd/kgM and rod-average fission gas releases of ∼0.3 to 11%. Measured concentrations of nuclides agreed within ∼10 to 15% of the predicted amounts. Radiochemistry of deposits on the cladding interior surface, fuel ceramography, beta/gamma autoradiog-raphy, and gamma scan data were consistent with fuel burnups and apparent operating temperatures. Microstructural observations also indicated increased porosity at the fuel center in high-release fuels and at the edge in high-burnup fuels. Cladding oxide thickness increased with exterior cladding temperatures as did the trend in hydriding.

Axial Fission Gas Transport in Nuclear Fuel Rods

2009 ◽  
Vol 283-286 ◽  
pp. 262-267
Author(s):  
M.T. del Barrio ◽  
Luisen E. Herranz
Keyword(s):  
Nuclear Fuel ◽  
Gas Transport ◽  
Fuel Pellet ◽  
Gas Release ◽  
Gas Mixing ◽  
Fuel Performance ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fission Gas ◽  
Fission Gas Release

Fission of fissile uranium or plutonium nucleus in nuclear fuel results in fission products. A small fraction of them are volatile and can migrate under the effect of concentration gradients to the grain boundaries of the fuel pellet. Eventually, some fission gases are released to the rod void volumes by a thermally activated process. Local transients of power generation could distort even further the already non-uniform axial power and fission gas concentration profiles in fuel rods. Most of the current fuel rod performance codes neglects these gradients and the resulting axial fission gas transport (i.e., gas mixing is considered instantaneous). Experimental evidences, however, highlight axial gas mixing as a real time-dependent process. The thermal feedback between fission gas release, gap composition and fuel temperature, make the “prompt mixing assumption” in fuel performance codes a key point to investigate due to its potential safety implications. This paper discusses the possible scenarios where axial transport can become significant. Once the scenarios are well characterized, the available database is explored and the reported models are reviewed to highlight their major advantages and shortcomings. The convection-diffusion approach is adopted to simulate the axial transport by decoupling both motion mechanisms (i.e., convection transport assumed to be instantaneous) and a stand-alone code has been developed. By using this code together with FRAPCON-3, a prospective calculation of the potential impact of axial mixing is conducted. The results show that under specific but feasible conditions, the assumption of “prompt axial mixing” could result in temperature underestimates for long periods of time. Given the coupling between fuel rod thermal state and fission gas release to the gap, fuel performance codes predictions could deviate non-conservatively. This work is framed within the CSN-CIEMAT agreement on “Thermo-Mechanical Behaviour of the Nuclear Fuel at High Burnup”.

Spatial Variations in the Mechanical Properties of the Porcine Thoracic Aorta

2011 ◽  
Author(s):  
Jungsil Kim ◽  
Seungik Baek
Keyword(s):  
Mechanical Properties ◽  
Vascular Disease ◽  
Thoracic Aorta ◽  
Aortic Wall ◽  
Vascular Tissue ◽  
Experimental Studies ◽  
Wall Stress ◽  
Clinical Decision ◽  
Arterial Tree

Characterization of the mechanical properties of a blood vessel is essential in understanding the progression of a vascular disease and for computational studies of vascular adaptation. For example, stiffness of vascular tissue is one of the major indicators to diagnose the vascular disease and make a clinical decision. Although previous studies reported the heterogeneity of the mechanical properties of arterial wall along the arterial tree [2], little was taken account for its circumferential variations. With the lack of experimental studies for investigating the circumferential variation, the aortic wall is typically assumed to have uniform deformation. Our previous study, however, has observed that there are circumferential variations in aortic wall stress and stiffness [1]. In addition to our previous study, we investigate further regional variations of the porcine thoracic aorta in both circumferential and longitudinal directions during the inflation test. Hence, we additionally test the distal thoracic aorta at each anterior and posterior side, respectively, and compare with the proximal thoracic aorta.

The Development of Alumina Forming Austenitic Alloy for Core Application in Advanced Nuclear Reactors

2020 ◽  
Vol 999 ◽  
pp. 72-80
Author(s):  
Zhang Jian Zhou ◽  
Ling Zhi Chen ◽  
Yuan Gao ◽  
Qi Wang
Keyword(s):  
Mechanical Properties ◽  
High Temperature ◽  
Room Temperature ◽  
High Efficiency ◽  
Nuclear Reactors ◽  
Potential Applications ◽  
Core Components ◽  
High Burnup ◽  
Corrosive Environments

The development of materials for core components which can serve in high temperature corrosive environments for a long service time is crucial to realize high efficiency and high-burnup operation of advanced nuclear reactors. Alumina forming austenitic (AFA) alloy is a kind of promising materials with improved corrosion resistance as well as strength at elevated temperature. The progress on the composition design and characterization of AFA alloys are reviewed in this work for evaluation their potential applications in advanced nuclear reactors. AFA alloys without the addition of carbon have been fabricated. Microstructures were observed by SEM and TEM. Mechanical properties were measured at room temperature and high temperature.

Microstructural and fractographic characterization of B4C-Al cermets tested under dynamic and static loading

1989 ◽  
Vol 47 ◽  
pp. 562-563
Author(s):  
Gyeung Ho Kim ◽  
Mehmet Sarikaya ◽  
D. L. Milius ◽  
I. A. Aksay
Keyword(s):  
Thin Film ◽  
Mechanical Properties ◽  
Fracture Toughness ◽  
Static Loading ◽  
Carbon Deposition ◽  
Full Density ◽  
Unique Combination ◽  
Strong Component ◽  
Reaction Products

Cermets are designed to optimize the mechanical properties of ceramics (hard and strong component) and metals (ductile and tough component) into one system. However, the processing of such systems is a problem in obtaining fully dense composite without deleterious reaction products. In the lightweight (2.65 g/cc) B4C-Al cermet, many of the processing problems have been circumvented. It is now possible to process fully dense B4C-Al cermet with tailored microstructures and achieve unique combination of mechanical properties (fracture strength of over 600 MPa and fracture toughness of 12 MPa-m1/2). In this paper, microstructure and fractography of B4C-Al cermets, tested under dynamic and static loading conditions, are described.The cermet is prepared by infiltration of Al at 1150°C into partially sintered B4C compact under vacuum to full density. Fracture surface replicas were prepared by using cellulose acetate and thin-film carbon deposition. Samples were observed with a Philips 3000 at 100 kV.

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