Analysis of UO2-BeO Fuel Performance During Normal Conditions and RIA

2018 ◽  
Author(s):  
Yanan He ◽  
Yingwei Wu ◽  
Shihuai Wang ◽  
Bowen Qiu ◽  
G. H. Su
Keyword(s):  
Thermal Conductivity ◽  
Thermal Properties ◽  
Gas Release ◽  
Fuel Performance ◽  
Loss Of Coolant Accident ◽  
Fission Gas ◽  
Composite Fuel ◽  
And Behaviors ◽  
Water Reactors ◽  
Fission Gas Release

UO2-BeO composite fuel may enable Light Water Reactors (LWRs) to have better safety due to its higher thermal conductivity. Much work have been done on the analysis of UO2-BeO fuel performance during LWRs steady state and Loss of Coolant Accident (LOCA) conditions using hypothetical thermal properties and behaviors models, leading to much uncertainty of the results. In this paper, firstly, fuel swelling and densification models for UO2-BeO fuel were developed based on Halden experiment data. Secondly, UO2-BeO fuel thermal properties and behaviors models have been coded in FRAPCON4.0 and FRAPTRAN2.0 after an evaluation of their applicability to UO2-BeO performance simulation. Then, both UO2-BeO composite fuel and traditional UO2 fuel performance during normal conditions and RIA were done in this paper by modified version of FRAPCON4.0 and FRAPTRAN2.0. Finally, comparisons between UO2-BeO and UO2 performance were conducted. The results shows that the peaking temperature of fuel can be reduced about 200K and 150K during normal conditions and RIA by adopting UO2-BeO, respectively. At the same time, the onset of pellet-cladding mechanic interaction (PCMI) can be delayed about 100days during normal conditions and the weakened PCMI effect can be expected during reactivity insertion accidents (RIA) due to the lower thermal expansion coefficient and temperature distribution for UO2-BeO composite fuel. Also, enthalpy stored in UO2-BeO fuel is reduced about 1/5 compared with that of UO2. However, fission gas release ration of UO2-BeO was a bit larger than that of UO2 due to its higher average burnup. And, further experiments stilled required to gain data for UO2-BeO during high burnup, like possibly reduced thermal conductivity and fission gas release threshold.

A Method to Evaluate Fission Gas Release During Irradiation Testing of Spherical Fuel

2008 ◽  
Author(s):  
Hanno van der Merwe ◽  
Johan Venter
Keyword(s):  
Fission Products ◽  
Gas Release ◽  
Fuel Performance ◽  
Coated Particle ◽  
Fission Gas ◽  
Online Measurements ◽  
Materials Used ◽  
Fission Gas Release

The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of Krypton and Xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6 and EU1bis.

Multiphysics Modeling of Thorium-Based (Th, U)O2 and (Th, Pu)O2 Fuel Performance in a Light Water Reactor

2018 ◽  
Author(s):  
Rong Liu ◽  
Jie-Jin Cai ◽  
Wen-Zhong Zhou ◽  
Ye Wang
Keyword(s):  
Fuel Burnup ◽  
Light Water Reactor ◽  
Gas Release ◽  
Fuel Performance ◽  
Light Water ◽  
Water Reactor ◽  
Fission Gas ◽  
Fission Gas Release ◽  
Centerline Temperature ◽  
Plenum Pressure

ThO2 has been considered as a possible replacement for UO2 fuel for future generation of nuclear reactors, and thorium-based mixed oxide (Th-MOX) fuel performance in a light water reactor was investigated due to better neutronics properties and proliferation resistance compared to conventional UO2 fuel. In this study, the thermal, mechanical properties of Th0.923U0.077O2 and Th0.923Pu0.077O2 fuel were reviewed with updated properties and compared with UO2 fuel, and the corresponding fuel performance in a light water reactor under normal operation conditions were also analyzed and compared by using CAMPUS code. The Th0.923U0.077O2 fuel were found to decrease the fuel centerline temperature, while Th0.923Pu0.077O2 fuel was found to have a bit higher fuel centerline temperature than UO2 fuel at the beginning of fuel burnup, and then much lower fuel centerline than UO2 fuel at high fuel burnup. The Th0.923U0.077O2 fuel was found to have lowest fuel centerline temperature, fission gas release and plenum pressure. While the Th0.923Pu0.077O2 fuel was found to have earliest gap closure time with much less fission gas release and much lower plenum pressure compared to UO2 fuel. So the fuel performance could be expected to be improved by applying Th0.923U0.077O2 and Th0.923Pu0.077O2 fuel.

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”.

scholarly journals Using laser remote heating to simulate extreme thermal heat loads on nuclear fuels

2020 ◽  
Vol 225 ◽  
pp. 08002
Author(s):  
T. Vidal ◽  
L. Gallais ◽  
J. Faucheux ◽  
H. Capdevila ◽  
Y. Pontillon
Keyword(s):  
Structural Changes ◽  
Temporal Dynamics ◽  
Experimental Chamber ◽  
Materials Testing ◽  
Gas Release ◽  
Absolute Level ◽  
Loss Of Coolant Accident ◽  
Fission Gas ◽  
Spatio Temporal ◽  
Fission Gas Release

Up to now, predicting accurately the Fission Gas Release (FGR) from high burn up UO2 and/or MOX (Mixed Oxide) fuels at off-normal conditions, such as power transient, reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA), is still a significant and very challenging task. For this purpose, different R&D programs have been carried out in France, as well as in other countries. This has been done with a specific emphasis on mechanisms which promote the FGR under accidental conditions. These studies can be performed thanks to dedicated integral experiments conducted in-pile (i.e. in Materials Testing Reactor) with the corresponding cost and constraints, or at the laboratory scale with annealing tests which allow to be representative of specific parameters (thermal history for instance). During these annealing tests under well-known conditions (temperature, atmosphere), both the absolute level and the time dependence of the released gases should be monitored, together with the corresponding fuel micro-structural changes, since experimental knowledge of fission gas release alone is not efficient enough. This approach requires more and more accurate on-line measurements. This corresponds to the driving force of the present work. In this contribution, we will present our progress in developing an experimental platform that can submit nuclear fuel and cladding samples to annealing tests involving very high temperatures (up to 2500°C) and very fast temperature ramp (up to thousands of °C/s) with controlled thermal gradients and temporal dynamics. This new platform implements innovative instrumentation, such as optical diagnostics to measure fuel fragmentation kinetics and infrared pyrometry for temperature monitoring. This experiment is based on a high-power laser (1.5kW) coupled to an experimental chamber with controlled atmosphere (Ar, N2, or vacuum) and specific optical components. Based on the spatial beam profile and temporal power function of the laser, it is possible which such a system to produce complex spatio-temporal temperature gradients, relevant for addressing different research needs. It provides access to extreme conditions that are very difficult to reach with other means. Particularly, one of main objectives of this work is to investigate conditions of Reactivity Initiated Accident (RIA). The first experiments performed on inactive materials, non-irradiated uranium dioxide, is presented in order to highlight the capabilities of this technique.

A Method to Evaluate Fission Gas Release During Irradiation Testing of Spherical Fuel

2009 ◽  
Vol 131 (5) ◽  
Author(s):  
Hanno van der Merwe ◽  
Johan Venter
Keyword(s):  
Fission Products ◽  
Gas Release ◽  
Fuel Performance ◽  
Coated Particle ◽  
Fission Gas ◽  
Online Measurements ◽  
Materials Used ◽  
Fission Gas Release

The evaluation of fission gas release from spherical fuel during irradiation testing is critical to understand expected fuel performance under real reactor conditions. Online measurements of krypton and xenon fission products explain coated particle performance and contributions from graphitic matrix materials used in fuel manufacture and irradiation rig materials. Methods that are being developed to accurately evaluate fission gas release are described here together with examples of evaluations performed on irradiation tests HFR-K5, -K6, and EU1bis.

Assessment of TRANSURANUS Fuel Performance Code Against Studsvik Inter-Ramp BWR Database

2009 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Paul Van Uffelen ◽  
Francesco Oriolo ◽  
Francesco D’Auria
Keyword(s):  
Experimental Data ◽  
Model Development ◽  
Normal Operation ◽  
Design Parameters ◽  
Gas Release ◽  
Fuel Performance ◽  
Data Set ◽  
Starting Point ◽  
Fission Gas ◽  
Fission Gas Release

The fuel matrix and the cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well. Investigations of fuel behavior are carried out in close connection with experimental research operation feedback and computational analyses. In this connection, OECD NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation - International Fuel Performance Experiments (IFPE) database”, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for model development and code validation. This database includes the data set of the Studsvik Inter-Ramp BWR Project. The objectives of the project are to establish the failure-safe operating limits and the failure mechanism and associated phenomena, during power ramp tests, by varying the design parameters (i.e. cladding heat treatment, gap thickness and fuel density). The experimental data are used for the assessment of the Fission Gas Release (FGR) models implemented in the TRANSURANUS code versions “v1m1j07” and “v1m1j08”. The starting point of the activity is the availability of a “new” transient fission gas release model, the “TFGR model”, specifically implemented in the last code version, to cover power ramp conditions. The paper presents the complete set of simulations of all twenty rods irradiated in the R2 research reactor and the corresponding comparisons with the experimental data. Sensitivity calculations are also performed to address the influence of geometric parameters and the choice of the different code options, relevant to model the FGR, on results.

Multiphysics Modeling of Fabrication Methods’ Effect on UO2-BeO Composite Fuels’ Performance

2018 ◽  
Author(s):  
Wei Zhou ◽  
Wenzhong Zhou
Keyword(s):  
Thermal Conductivity ◽  
Volume Fraction ◽  
Particle Volume Fraction ◽  
Particle Volume ◽  
Fuel Performance ◽  
Micro Cracks ◽  
Composite Fuel ◽  
Johnson Model ◽  
Fission Gas Release ◽  
Good Agreement

UO2-BeO is one of the most promising accident tolerant nuclear fuels due to its excellent thermal conductivity compared to pure UO2 fuel. Two different UO2-BeO fabrication methods have demonstrated the capability to fabricate enhanced thermal conductivity UO2-BeO composite fuels and improve fuel performance. In one method, BeO is continuously distributed around UO2 grains while BeO is dispersed in the UO2 phase in the other method. In the former type, BeO is considered as matrix and UO2 as dispersed particle. However, in the later type, BeO is considered as dispersed particle and UO2 as matrix. To calculate the thermal conductivity of UO2-BeO composite, Hasselman-Johnson model has been applied, which shows good agreement with experimental data. In this model, it includes the influence of thermal conductivity of matrix and particle, volume fraction of particle, radius of particle and the interfacial thermal conductivity between matrix and particle. To balance the improvement of thermal conductivity and enrichment of UO2, a UO2-BeO composite fuel with 10% volume fraction of BeO has been chosen. Besides, the grain size of matrix should be noticeable smaller than particle to relax thermal stress which may cause micro-cracks and destruction, leading to the grain sizes of UO2 in two types being distinctive, and resulting in very big effect on fission gas release. In the paper the thermal conductivity has been intensively studied as well as fuel performance in two different types of UO2-BeO fuels, and the two fabrication methods have also been compared to assess their applications in commercial reactors.

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