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.

Effects of matrix composition and sample preparation on instant release fractions from high burnup nuclear fuel

2014 ◽  
Vol 1665 ◽  
pp. 261-266 ◽  
Author(s):  
O. Roth ◽  
J. Low ◽  
K. Spahiu
Keyword(s):  
Sample Preparation ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Matrix Composition ◽  
Spent Fuel ◽  
Preparation Methods ◽  
Rapid Release ◽  
Leaching Solution ◽  
First Contact

ABSTRACTThe rapid release of fission products segregated either to the gap between the fuel and the cladding or to the UO2 grain boundaries from spent nuclear fuel in contact with water (often referred to as the instant release fraction - IRF) is of interest for the safety assessment of geological repositories for spent fuel due to the potential dose contribution. In September 2012 a study was initiated with the aim of comparing the instant release behavior of fuels with and without additives/dopants. Preliminary results from this (ongoing) study indicate that the release of uranium during the first contact periods was higher than during the tests with fuel segments, even though the fuel was cut open recently [1]. This could be due to the sample preparation method which included axial cutting of the cladding in order to remove the fuel fragments used in the study. In the present work, leaching data from both studies are presented and the releases are discussed comparing the two sample preparation methods and considering the effect of matrix composition. The leaching studies have been performed in air using 10 mM NaCl + 2 mM NaHCO3 as leaching solution.

Inventories of Iodine-129 and Cesium-137 in the Gaps and Grain Boundaries of LWR Spent Fuels

1999 ◽  
Vol 556 ◽  
Author(s):  
W. J. Gray
Keyword(s):  
Grain Boundary ◽  
Spent Nuclear Fuel ◽  
Yucca Mountain ◽  
Light Water Reactor ◽  
Gas Release ◽  
Spent Fuel ◽  
Fission Gas ◽  
Geologic Disposal ◽  
Cesium 137 ◽  
Fission Gas Release

AbstractPerformance assessment calculations that support geologic disposal of spent nuclear fuel in a potential repository at Yucca Mountain, Nevada, are based in part on the assumption that 2% of the total inventories of 135Cs, 129I, and 99Tc are located in the gap and grain-boundary regions where they could dissolve rapidly if the spent fuel were to be contacted by groundwater. Actual measured values reported here for a few light-water reactor (LWR) spent fuels show that the combined gap and grain-boundary inventories of 129I approximately equaled the fission-gas release fractions. For 137Cs, the combined gap and grain-boundary inventories were approximately one third of the fission-gas release fractions. These measured values can be used to replace the 2% estimate and thus reduce the uncertainties in the calculations.

Distribution of Palladium During Spent Fuel Reprocessing

2003 ◽  
Author(s):  
Yu. Pokhitonov ◽  
V. Romanovski ◽  
P. Rance
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Platinum Group Metals ◽  
Spent Fuel ◽  
Platinum Metals ◽  
Spent Fuel Reprocessing ◽  
Platinum Group ◽  
Fuel Reprocessing

The principal purpose of spent fuel reprocessing consists in the recovery of the uranium and plutonium and the separation of fission products so as to allow re-use of fissile and fertile isotopes and facilitate disposal of waste elements. Amongst the fission products present in spent nuclear fuel of Nuclear Power Plants (NPPs,) there are considerable quantities of platinum group metals (PGMs): ruthenium, rhodium and palladium. Given current predictions for nuclear power generation, it is predicted that the quantities of palladium to be accumulated by the middle of this century will be comparable with those of the natural sources, and the quantities of rhodium in spent nuclear fuel may even exceed those in natural sources. These facts allow one to consider spent nuclear fuel generated by NPPs as a potential source for creation of a strategic stock of platinum group metals. Despite of a rather strong prediction of growth of palladium consumption, demand for “reactor” palladium in industry should not be expected because it contains a long-lived radioactive isotope 107Pd (half-life 6,5·105 years) and will thus be radioactive for a very considerable period, which, naturally, restricts its possible applications. It is presently difficult to predict all the areas for potential use of “reactor” palladium in the future, but one can envisage that the use of palladium in radwaste reprocessing technology (e.g. immobilization of iodine-129 and trans-plutonium elements) and in the hydrogen energy cycle may play a decisive role in developing the demand for this metal. Realization of platinum metals recovery operation before HLW vitrification will also have one further benefit, namely to simplify the vitrification process, because platinum group metals may in certain circumstances have adverse effects on the vitrification process. The paper will report data on platinum metals (PGM) distribution in spent fuel reprocessing products and the different alternatives of palladium separation flowsheets from HLW are presented. It is shown, that spent fuel dissolution conditions can affect the palladium distribution between solution and insoluble precipitates. The most important factors, which determine the composition and the yield of residues resulting from fuel dissolution, are the temperature and acid concentration. Apparently, a careful selection of fuel dissolution process parameters would make it possible to direct the main part of palladium to the 1st cycle raffinate together with the other fission products. In the authors’ opinion, the development of an efficient technology for palladium recovery requires the conception of a suitable flow-sheet and the choice of optimal regimes of “reactor” palladium recovery concurrently with the resolution of the problem of HLW partitioning when using the same facilities.

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.

RN Fractional Release of High Burn-Up Fuel: Effect of HBS and Estimation of Accessible Grain Boundary

2008 ◽  
Vol 1107 ◽  
Author(s):  
F. Clarens ◽  
D. Serrano-Purroy ◽  
A. Martínez-Esparza ◽  
D. Wegen ◽  
E. Gonzalez-Robles ◽  
...  
Keyword(s):  
Grain Boundary ◽  
Correction Factor ◽  
Core Sample ◽  
Gas Release ◽  
Spent Fuel ◽  
Fission Gas ◽  
Leaching Experiments ◽  
Fission Gas Release ◽  
Fractional Release ◽  
Number Of Particles

AbstractThe so-called Instant Release Fraction (IRF) is considered to govern the dose released from Spent Fuel repositories. Often, IRF calculations are based on estimations of fractions of inventory release based in fission gas release [1]. The IRF definition includes the inventory located within the Gap although a conservative approach also includes both the Grain Boundary (GB) and the pores of restructured HBS inventories.A correction factor to estimate the fraction of Grain Boundary accessible for leaching has been determined and applied to spent fuel static leaching experiments carried out in the ITU Hot Cell facilities [2]. Experimental work focuses especially on the different properties of both the external rim area (containing the High Burn-up Structure (HBS)) and the internal area, to which we will refer as Out and Core sample, respectively. Maximal release will correspond to an extrapolation to simulate that all grain boundaries or pores are open and in contact with solution.The correction factor has been determined from SEM studies taking into account the number of particles with HBS in Out sample, the porosity of HBS particles, and the amount of transgranular fractures during sample preparation.

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