Reduction of fission gas release at extended burnups in light water reactors by decreasing fuel temperature

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.

Download Full-text

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

Journal of Nuclear Materials ◽

10.1016/s0022-3115(96)00731-3 ◽

1997 ◽

Vol 244 (2) ◽

pp. 131-140 ◽

Cited By ~ 36

Author(s):

Toshiaki Kogai

Keyword(s):

Light Water Reactor ◽

Gas Release ◽

Light Water ◽

Water Reactor ◽

Fission Gas ◽

Fission Gas Release

Download Full-text

Multiphysics Modeling of Thorium-Based (Th, U)O2 and (Th, Pu)O2 Fuel Performance in a Light Water Reactor

Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory ◽

10.1115/icone26-81237 ◽

2018 ◽

Cited By ~ 1

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.

Download Full-text

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

Nuclear Technology ◽

10.13182/nt93-a34818 ◽

1993 ◽

Vol 102 (2) ◽

pp. 210-231 ◽

Cited By ~ 13

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

Download Full-text

IMPLEMENTATION AND VALIDATION OF A FISSION GAS RELEASE MODEL FOR CTFFUEL USING THE NEA/OECD IFPE DATABASE

EPJ Web of Conferences ◽

10.1051/epjconf/202124710018 ◽

2021 ◽

Vol 247 ◽

pp. 10018

Author(s):

A. Abarca ◽

M. Avramova ◽

K. Ivanov

Keyword(s):

Heat Transfer ◽

Material Behavior ◽

Gas Release ◽

Fuel Temperature ◽

Conductivity Equation ◽

Neutron Kinetics ◽

Fuel Rod ◽

Fission Gas ◽

Release Model ◽

Fission Gas Release

The nuclear reactors themselves are complex systems whose responses are driven by interactions between different physics phenomena within the reactor core. Traditionally, the different physics phenomena have been analyzed separately and its interaction considered via boundary conditions or closure models. However, in parallel with the development of computational technology, multi-physics coupled simulations are being used to obtain accurate predictions thanks to the consideration of the feedback effects on the fly (on-line). In the nuclear systems the fuel temperature is an important feedback parameter used to obtain the nuclear cross sections at given conditions by the neutron kinetics codes. An accurate prediction of temperature profile within the fuel rod involve several physics such as neutron kinetics, mechanics, material behavior and properties, heat transfer, thermal-hydraulics, and even chemistry. The pellet to clad gap conductance is possibly the most important source of uncertainty in the solution of conductivity equation in the fuel rod and the fuel temperature prediction. The gap conductance depends on two effects: the pellet to gap distance and the conductivity of the gas species that fill the gap. In this research work, the authors are focused on improving of the prediction of the gap gas conductivity in CTFFuel by implementing a fission gas release model in the code. The objective of this contribution is the implementation of a transient fission gas release model in CTFFuel and its validation using the experimental data available in the OECD/NEA International Fuel Performance Experiments (IFPE) database. CTFFuel is an isolated fuel heat transfer capability within the framework of CTF code, the state-of-the-art version of the Coolant Boiling in Rod Arrays Code – Two-Fluid (COBRA-TF) sub-channel thermal-hydraulic code. The code is being jointly developed by North Carolina State University (NCSU) and Oak Ridge National Laboratory (ORNL) within the US Department of Energy (DOE) Consortium for Advanced Simulation of LWRs (CASL).

Download Full-text

Analysis of UO2-BeO Fuel Performance During Normal Conditions and RIA

Volume 9: Student Paper Competition ◽

10.1115/icone26-82076 ◽

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.

Download Full-text