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.

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”.

Method for Analyzing Fission Gas Release in Fuel Rods Based on Gamma-Ray Measurements of Short-Lived Fission Products

2013 ◽  
Vol 184 (1) ◽  
pp. 96-106 ◽  
Author(s):  
Scott Holcombe ◽  
Staffan Jacobsson Svärd ◽  
Knut Eitrheim ◽  
Lars Hallstadius ◽  
Christofer Willman
Keyword(s):  
Gamma Ray ◽  
Fission Products ◽  
Gas Release ◽  
Fuel Rods ◽  
Fission Gas ◽  
Fission Gas Release

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.

Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

2012 ◽  
Vol 420 (1-3) ◽  
pp. 54-62 ◽  
Author(s):  
L. Johnson ◽  
I. Günther-Leopold ◽  
J. Kobler Waldis ◽  
H.P. Linder ◽  
J. Low ◽  
...  
Keyword(s):  
Fission Products ◽  
Experimental Results ◽  
Gas Release ◽  
Fission Gas ◽  
Fission Gas Release

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.

Effects of matrix composition on instant release fractions from high burn-up nuclear fuel

2013 ◽  
Vol 1518 ◽  
pp. 145-150 ◽  
Author(s):  
Olivia Roth ◽  
Jeanett Low ◽  
Michael Granfors ◽  
Kastriot Spahiu
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Matrix Composition ◽  
Initial Contact ◽  
Gas Release ◽  
Spent Fuel ◽  
Rapid Release ◽  
Fission Gas ◽  
Fission Gas Release

ABSTRACTThe release of radionuclides from spent nuclear fuel in contact with water is controlled by two processes – the dissolution of the UO2 grains and the rapid release of fission products segregated either to the gap between the fuel and the cladding or to the UO2 grain boundaries. The rapid release is often referred to as the Instant Release Fraction (IRF) and is of interest for the safety assessment of geological repositories for spent fuel due to the potential dose contribution.Previous studies have shown that the instant release fraction can be correlated to the fission gas release (FGR) from the spent fuel. Studies comparing results from samples in the form of pellets, fragments, powders and a fuel rodlet have shown that the sample preparation has a significant impact on the instant release, indicating that the differentiation between gap release and grain boundary release should be further explored.Today, there are trends towards power uprates, longer fuel cycles and increasing burn-up putting additional requirements on the nuclear fuel. These requirements are met by the development of new fuel types, such as UO2 fuels containing dopants or additives. The additives and dopants affect fuel properties such as grain size and fission gas release. In the present study we have performed experimental leaching studies using two high burnup fuels with and without additives/dopants and compared the fuel types with respect to their instant release behavior. The results of the leaching of the samples for the 3 initial contact periods; 1, 7 and 23 days are reported here.

Sign in / Sign up

Export Citation Format

Share Document