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.

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.

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

Pellet-Cladding Interaction of LMFBR Fuel Elements at Unsteady State: An Introduction to Computer Code ISUNE-5

1981 ◽  
Vol 103 (4) ◽  
pp. 627-636 ◽  
Author(s):  
B. M. Ma
Keyword(s):  
Reactor Core ◽  
Fatigue Cracking ◽  
Computer Code ◽  
Fuel Burnup ◽  
Fuel Pellet ◽  
Gas Release ◽  
Unsteady State ◽  
Fission Gas ◽  
Fuel Elements ◽  
Fission Gas Release

The fuel pellet-cladding interaction (PCI) of liquid-metal fast breeder reactor (LMFBR) fuel elements or fuel rods at unsteady state is analyzed and discussed based on experimental results. In the analyses, the heat generation, fuel restructuring, temperature distribution, gap conductance, irradiation swelling, irradiation creep, fuel burnup, fission gas release, fuel pellet cracking, crack healing, cladding cracking, yield failure and fracture failure of the fuel elements are taken into consideration. To improve the sintered (U,Pu)O2 fuel performance and reactor core safety at high temperature and fuel burnup, it is desirable to (a) increase and maintain the ductility of cladding material, (b) provide sufficient gap thickness and plenum space for accommodating fission gas release, (c) keep ramps-power increase rate slow and gentle, and (d) reduce the intensity and frequency of transient PCI in order to avoid intense stress fatigue cracking (SFC) and stress corrosion cracking (SCC) due to fission product compounds CsI, CdI2, Cs2Te, etc. at the inner cladding surface of the fuel elements during PCI.

Multiphysics coupled modeling of light water reactor fuel performance

2016 ◽  
Vol 91 ◽  
pp. 38-48 ◽  
Author(s):  
Rong Liu ◽  
Andrew Prudil ◽  
Wenzhong Zhou ◽  
Paul K. Chan
Keyword(s):  
Reactor Fuel ◽  
Light Water Reactor ◽  
Coupled Modeling ◽  
Fuel Performance ◽  
Light Water ◽  
Water Reactor

scholarly journals Planning support document for the EPRI Light Water Reactor Fuel Performance program

1977 ◽  
Author(s):  
J. T.A. Roberts ◽  
F. E. Gelhaus ◽  
H. Ocken ◽  
S. T. Oldberg ◽  
G. R. Thomas
Keyword(s):  
Reactor Fuel ◽  
Light Water Reactor ◽  
Fuel Performance ◽  
Light Water ◽  
Water Reactor ◽  
Planning Support

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.

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