• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.155-162
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• 2009

Since the 90's, EDF and AREVA-NP have irradiated, up to very high burnups, lead assemblies housing $M5^{(R)}$ cladded fuels. Post-irradiation examination of high burnup $UO_2$ pellets show an increase in the fission-gas release rate, an increase in fuel swelling, and formation of fission-gas bubbles throughout the pellets. Xenon abundances were quantified, and phenomena leading to this bubble formation were identified. All examinations provided valuable data on the complex state of the fuel during irradiation. They show the good behavior of these fuels, exhibiting various microstructures at very high burnups, none of which is likely to lead to problems during irradiation.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.169-182
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• 2014

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2816-2829
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• 2021

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3133-3150
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• 2021

This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.299-308
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• 2005

The behavior of fission gas in all major types of nuclear fuel has been reviewed with an emphasis on more recently discovered aspects. It is proposed that the behavior of fission gas can be classified in a number of characteristic types that occur at a high or low operating temperature, and/or at high or low fissile burnup. The crystal structure and microstructure of the various fuels are the determinant factors in the proposed classification scheme. Three types of behavior, characterized by anisotropic $\alpha$-U, high temperature metallic $\gamma$-U, and cubic ceramics, are well-known and have been extensively studied in the literature. Less widely known are two equally typical low temperature kinds: one associated with fission induced grain refinement and the other with fission induced amorphization. Grain refinement is seen in crystalline fuel irradiated to high burnup at low temperatures, whereas breakaway swelling is observed in amorphous fuel containing sufficient excess free-volume. Amorphous fuel, however, shows stable swelling if insufficient excess free-volume is available during irradiation.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.79-88
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• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.66-66
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.251-251
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• 1998

• Nuclear Engineering and Technology
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• v.37 no.4
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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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• v.45 no.4
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• pp.421-426
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• 2013

The need for extended storage of used nuclear fuel is increasing globally as disposition schedules for used fuel are pushed further into the future. This is creating a situation where dry storage of used fuel may need to be extended beyond normal regulatory licensing periods. While it is generally accepted that used fuel in dry storage will remain in a safe condition, there is little data that demonstrate used fuel performance in dry storage environments for long periods of time. This is especially true for high burnup used fuel. This paper discusses a technical approach that defines a process that develops the technical basis for demonstrating the safety of used fuel over extended periods of time.