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.

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.

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.

scholarly journals Initial Results from Dissolution Testing of Various Air-Oxidized Spent Fuels

1993 ◽  
Vol 333 ◽  
Author(s):  
W. J. Gray ◽  
L. E. Thomas
Keyword(s):  
Light Water Reactor ◽  
Dissolution Testing ◽  
Gas Release ◽  
Test Results ◽  
Light Water ◽  
Water Reactor ◽  
Dissolution Tests ◽  
Fission Gas ◽  
Fission Gas Release ◽  
Initial Results

ABSTRACTFlowthrough dissolution tests have been conducted on two different light-water-reactor spent fuels oxidized to U4O9+x or U3O8. Oxidation had a bigger impact on the dissolution of U and a smaller impact on the dissolution of Tc from the fuel with higher burnup and higher fission gas release. Possible reasons for the observed differences in test results are discussed, but clarification awaits results from tests on other fuels, which are in progress.

Powder Leaching Study for Grain Boundary Inventory of Two High Burnup Fuels

MRS Advances ◽  
2019 ◽  
Vol 4 (17-18) ◽  
pp. 981-986
Author(s):  
Alexandre Barreiro Fidalgo ◽  
Olivia Roth ◽  
Anders Puranen ◽  
Lena Z. Evins ◽  
Kastriot Spahiu ◽  
...  
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Spent Fuel ◽  
Short Period ◽  
Fission Gas ◽  
Leaching Solution ◽  
Significant Release ◽  
Leaching Experiments ◽  
High Burnup ◽  
Fission Gas Release

ABSTRACTIn the context of safety assessment, the fraction of inventory that is expected to rapidly dissolve when water contacts the spent fuel is called the Instant Release Fraction (IRF). Conceptually, this fraction consists of radionuclides outside of the uranium dioxide matrix and therefore the fraction can be further divided into the radionuclides in the fuel/cladding gap and radionuclides in the grain boundaries. The relative importance of these two fractions is investigated here for two Swedish high burnup fuels through simultaneous grinding and leaching fuel fragments in simplified groundwater for a short period of time. The hypothesis is that this will expose grain boundaries to leaching solution and provide an estimate of the release of the grain boundary inventory upon contact with water. The studied fragments were used in previous leaching experiments and thus pre-washed to remove any pre-oxidized phases. The results showed a significant release of iodine, cesium and rubidium and to a lower extent molybdenum and technetium. The fraction of inventory in the aqueous phase of actinides and lanthanides was 1-2 orders of magnitude lower than for the elements associated to the IRF. Both fuels displayed a very similar behavior and no correlation as a function of burnup or fission gas release was found.

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.

Measurements of Grain-Boundary Inventories of 137Cs, 90Sr and 99Tc in Used Candu Fuel

1992 ◽  
Vol 294 ◽  
Author(s):  
S. Stroes-Gascoyne ◽  
J.C. Tait ◽  
R.J. Porth ◽  
J.L. Mcconnell ◽  
T.R. Barnsdale ◽  
...  
Keyword(s):  
Aqueous Solution ◽  
Grain Boundary ◽  
Grain Boundaries ◽  
Source Term ◽  
Slow Release ◽  
Gas Release ◽  
System A ◽  
Fission Gas ◽  
Fission Gas Release ◽  
Upper Boundary

ABSTRACTTwo methods were used to measure grain-boundary inventories of 137Cs, 90Sr and 99Tc in used CANDU fuel, to corroborate source term estimates based on a fission gas release code. Used fuels were partially oxidized at 200°C in air to overall compositions of UO2+x (0.15≤ × ≤0.25) to expose UO2 grain boundaries, followed by leaching in aqueous solution. Only a fraction (2 to 18%) of the calculated gap + grain-boundary inventories for 37Cs was released. This suggests that the calculations overestimate Cs release or that oxidation does not expose all grain boundaries, or that Cs release from grain boundaries is slow. Release of 90Sr (0.01 to 0.7%) agreed reasonably well with the source term estimates (0.001 to 0.3%). Release of 99Tc (0.3 to 1.5%) suggests that the source term estimate for the upper boundary of 99Tc release (25%) may be too high. A second technique involved leaching of crushed and size-fractionated used fuel in either a static or dynamic system. A direct one-to-one correlation between calculated and measured gap + grain-boundary inventories for 137Cs was found for low- and medium-power fuels.

Modeling the influence of bubble pressure on grain boundary separation and fission gas release

2014 ◽  
Vol 452 (1-3) ◽  
pp. 95-101 ◽  
Author(s):  
Pritam Chakraborty ◽  
Michael R. Tonks ◽  
Giovanni Pastore
Keyword(s):  
Grain Boundary ◽  
Gas Release ◽  
Boundary Separation ◽  
Bubble Pressure ◽  
Fission Gas ◽  
Fission Gas Release
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