Modelling of fission gas release and gaseous swelling of light water reactor fuels

1997 ◽  
Vol 244 (2) ◽  
pp. 131-140 ◽  
Author(s):  
Toshiaki Kogai
Keyword(s):  
Light Water Reactor ◽  
Gas Release ◽  
Light Water ◽  
Water Reactor ◽  
Fission Gas ◽  
Fission Gas Release

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.

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.

Evaluation of Fission Gas Release in High-Burnup Light Water Reactor Fuel Rods

1993 ◽  
Vol 102 (2) ◽  
pp. 210-231 ◽  
Author(s):  
John O. Barner ◽  
Mitchel E. Cunningham ◽  
Maxwell D. Freshley ◽  
Donald D. Lanning
Keyword(s):  
Reactor Fuel ◽  
Light Water Reactor ◽  
Gas Release ◽  
Fuel Rods ◽  
Light Water ◽  
Water Reactor ◽  
Fission Gas ◽  
High Burnup ◽  
Fission Gas Release

Analysis of Fission Gas Disposition in Light Water Reactor Steady-State Operation

1982 ◽  
Vol 58 (3) ◽  
pp. 492-510 ◽  
Author(s):  
Antonio Villalobos ◽  
A. R. Wazzan ◽  
D. Okrent
Keyword(s):  
Steady State ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Steady State Operation ◽  
Fission Gas

Local fission gas release and swelling in water reactor fuel during slow power transients

1985 ◽  
Vol 131 (2-3) ◽  
pp. 162-171 ◽  
Author(s):  
M. Mogensen ◽  
C.T. Walker ◽  
I.L.F. Ray ◽  
M. Coquerelle
Keyword(s):  
Reactor Fuel ◽  
Gas Release ◽  
Water Reactor ◽  
Fission Gas ◽  
Fission Gas Release

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.

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