Thin film samples: a new methodology for investigating the mechanisms of fission gas releases from nuclear fuel during a LOCA

ABSTRACTAccurately predicting fission gas releases (FGR) from high burn-up fuels during off-normal conditions, such as a loss-of-coolant accident (LOCA), is an important, major challenge. A significant step forward would be to identify and assess the basic mechanisms causing this FGR. A helpful way of better understanding these basic mechanisms is to separate these effects and to perform irradiations on materials simulating the nuclear fuel. Mesoporous or dense CeO2 and UO2 samples (with thin film geometry) were selected for these studies as materials representative of irradiated fuels. A basic mechanism to obtain a better understanding of FGR is described and a new methodology using thin film samples is developed to test the validity of this mechanism.

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Prediction of Ballooning and Burst for Nuclear Fuel Cladding with Anisotropic Creep Modeling During Loss of Coolant Accident (LOCA)

Nuclear Engineering and Technology ◽

10.1016/j.net.2021.04.020 ◽

2021 ◽

Author(s):

Jinsu Kim ◽

Jeong Whan Yoon ◽

Hyochan Kim ◽

Sung-Uk Lee

Keyword(s):

Nuclear Fuel ◽

Fuel Cladding ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Nuclear Fuel Cladding ◽

Anisotropic Creep

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A Study of Prediction Model Improvement for Air-Oxidation Breakaway in a Postulated Spent Nuclear Fuel Pool Complete Loss of Coolant Accident

Sustainability ◽

10.3390/su13031442 ◽

2021 ◽

Vol 13 (3) ◽

pp. 1442

Author(s):

Sanggil Park ◽

Jaeyoung Lee ◽

Min Bum Park

Keyword(s):

Kinetic Model ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Complete Loss ◽

Air Oxidation ◽

Loss Of Coolant Accident ◽

Spent Nuclear Fuel Pool ◽

Loss Of Coolant ◽

Model Derivation ◽

Model Improvement

The temperature of zirconium alloy cladding on the postulated spent nuclear fuel pool complete loss of coolant accident is abruptly increased at a certain time and the cladding is almost fully oxidized to weak ZrO2 in the air. This abrupt temperature escalation phenomenon induced by the air-oxidation breakaway is called a zirconium fire. Although an air-oxidation breakaway kinetic model correlated between time and temperature has been implemented in the MELCOR code, it is likely to bring about unexpected large errors because of many limitations of model derivation. This study suggests an improved time–temperature correlated kinetic model using the Johnson–Mehl equation. It is based on that the air-oxidation breakaway is initiated by the phase transformation from the tetragonal to monoclinic ZrO2 at the oxide–metal interface in the cladding. This new model equation is also evaluated with the Zry-4 air-oxidation literature data. This equation resulted in the almost similar air-oxidation breakaway timing to the actual experimental data at 800 °C. However, at 1000 °C, it showed an error of about 8 min. This could be inferred from the influence of the ZrN phase change due to the nitrogen existing in air.

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Rupture behaviour of nuclear fuel cladding during loss-of-coolant accident

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2016.07.022 ◽

2016 ◽

Vol 307 ◽

pp. 319-327 ◽

Cited By ~ 15

Author(s):

Siddharth Suman ◽

Mohd. Kaleem Khan ◽

Manabendra Pathak ◽

R.N. Singh ◽

J.K. Chakravartty

Keyword(s):

Nuclear Fuel ◽

Fuel Cladding ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Nuclear Fuel Cladding

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Modeling Axial Relocation of Fragmented Fuel During Loss of Coolant Conditions by Using ABAQUS

Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering ◽

10.1115/icone2020-16291 ◽

2020 ◽

Author(s):

Zehua Ma ◽

Koroush Shirvan ◽

Wei Li ◽

Yingwei Wu

Keyword(s):

Thermal Stresses ◽

Element Model ◽

Irradiation Creep ◽

Irradiation Hardening ◽

Fuel Pellets ◽

Loss Of Coolant Accident ◽

Fission Gas ◽

Loss Of Coolant ◽

Single Balloon ◽

Nuclear Fuel Pellets

Abstract In a light-water reactor, during normal operating condition, the UO2 nuclear fuel pellets undergo fragmentation primarily due to presence of thermal stresses, fission gas development and pellet-clad mechanical interaction. Under Loss of Coolant Accident (LOCA) conditions, a portion of fuel fragments can freely move downwards to the ballooning region due to the significant cladding deformation. The fuel relocation can localize the heat load and in turn accelerate the cladding balloon and burst process. Cladding burst is of great concern because of the potential for fuel dispersal into coolant and clad structural stability. In our work, we built up a finite element model considering cladding balloon, fuel relocation and its resultant thermal feedback during LOCA condition with ABAQUS. The clad balloon model includes phase transformation, swelling, thermal and irradiation creep, irradiation hardening and annealing and other important thermal-mechanical properties. The mass of relocation model was verified against the analytical cases of single balloon and twin balloons. The cladding balloon model combined with fuel thermal conductivity degradation was verified against fuel performance code, FRAPTRAN. Finally, with the evolution of pellet-cladding gap, the fuel mass relocation was calculated and compared against the IFA-650.4 transient test from the Halden reactor.

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Impact of hydrogen on rupture behaviour of Zircaloy-4 nuclear fuel cladding during loss-of-coolant accident: a novel observation of failure at multiple locations

Nuclear Engineering and Technology ◽

10.1016/j.net.2020.07.017 ◽

2020 ◽

Author(s):

Siddharth Suman

Keyword(s):

Nuclear Fuel ◽

Fuel Cladding ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Nuclear Fuel Cladding

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The VINON-LOCA test facility: exploring the LOCA phenomenology through an out-of-pile thermal sequence on irradiated pressurized fuel rod

EPJ Web of Conferences ◽

10.1051/epjconf/202125306002 ◽

2021 ◽

Vol 253 ◽

pp. 06002

Author(s):

B. Biard ◽

C. Colin ◽

S. Bernard ◽

V. Marty ◽

G. Volle ◽

...

Keyword(s):

Gamma Camera ◽

Test Facility ◽

Flow Meters ◽

Loss Of Coolant Accident ◽

Fission Gas ◽

Loss Of Coolant ◽

Post Test ◽

Set Up ◽

Fuel Fragmentation ◽

Fission Gas Release

Since the out-of-pile semi-integral tests performed at Studsvik in 2011 for the NRC [1] and the Halden Loss-Of-Coolant Accident (LOCA) test series IFA-650 [2], a major safety interest has raised for Fuel Fragmentation, Relocation and Dispersal (FFRD) during a LOCA sequence. In addition to the characteristics of the fuel ejected from the rod after the clad failure, the fuel behaviour before the clad failure is still to be investigated, especially its fragmentation and its possible relocation within the rod during the clad ballooning phase. Furthermore, the chronology and the sequencing of these phenomena is of particular interest. For this purpose, the VINON-LOCA program, lying in the framework of a trilateral agreement between EDF, Framatome and CEA, is aimed at performing Out-Of-Pile heating tests on irradiated repressurized fuel rods, reproducing a typical Loss Of Coolant Accident thermal sequence. The VINON-LOCA experimental set-up is located in the so-called VERDON lab of the LECA-STAR hot cell complex. This lab was dedicated to the VERDON-ISTP program [3]. The VINON-LOCA set-up is thus largely instrumented for addressing not only these FFRD topics, but also Fission Gas Release (FGR), combining both online measurement (gamma stations, gamma camera, acoustic sensor, pressure, temperatures, flow meters, microGC…), and preand post-test characterization (gamma scanning, tomography, metrology, fuel fragments weighing and sieving, gas analyses…). An extensive and substantial qualification campaign has been performed to validate the furnace design regarding the desired test conditions, and to qualify the instrumentation. Following some preliminary modelling and calculations, it has included tests on an out-of-cell twin mockup and tests on dummy inactive rods in the hot cell. This allowed achieving successfully the first experimental qualification test of the program end of 2019 on an irradiated UO2 fuel rodlet. A second irradiated experiment is planned with increased instrumentation capabilities, notably a 2D gamma camera for online fuel motion detection.

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Coupled Modeling and Simulation of Phase Transformation in Zircaloy-4 Fuel Cladding Under Loss-of-Coolant Accident Conditions

Frontiers in Energy Research ◽

10.3389/fenrg.2021.667419 ◽

2021 ◽

Vol 9 ◽

Author(s):

Wenjun Lu ◽

Libo Qian ◽

Wenzhong Zhou

Keyword(s):

Thermal Conductivity ◽

Phase Transformation ◽

Nuclear Fuel ◽

Transformation Model ◽

Fuel Cladding ◽

Fuel Rod ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Accident Conditions ◽

Enhanced Thermal Conductivity

Under loss-of-coolant conditions, the temperature on fuel cladding will increase rapidly (up to 1000–1500 K), which will not only cause a dramatic oxidation reaction of Zircaloy-4 and an increase in hydrogen concentration but also cause an allotropic phase transformation of Zircaloy-4 from hexagonal (α-pahse) to cubic (β-phase) crystal structure. As we all know, thermophysical properties have a close relationship with the microstructure of the material. Moreover, because of an important influence of the phase transformation on the creep resistance and the ductility of the fuel rod, studying the crystallographic phase transformation kinetics is pivotal for evaluating properties for fuel rod completeness. We coupled the phase transformation model together with the existing physical models for reactor fuel, gap, cladding, and coolant, based on the finite element analysis and simulation software COMSOL Multiphysics. The critical parameter for this transformation is the evolution of the volume fraction of the favored phase described by a function of time and temperature. Hence, we choose two different volume fractions (0 and 10%) of BeO for UO2-BeO enhanced thermal conductivity nuclear fuel and zircaloy cladding as objects of this study. In order to simulate loss-of-coolant accident conditions, five relevant parameters are studied, including the gap size between fuel and cladding, the temperature at the extremities of the fuel element, the coefficient of heat transfer, the linear power rate, and the coolant temperature, to see their influence on the behavior of phase transformation under non-isothermal conditions. The results show that the addition of 10vol%BeO in the UO2 fuel decreased the phase transformation effect a lot, and no significant phase transformation was observed in Zircaloy-4 cladding with UO2-BeO enhanced thermal conductivity nuclear fuel during existing loss-of-coolant accident conditions.

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