Fuel Performance Analysis Research Articles

Irradiation Behavior of Thoria-Urania Fuel in a PWR

A thoria-urania (ThO2-UO2) fuel performance analysis code was developed by adding performance models and material properties correlations for thoria-urania fuel to a UO2 fuel performance analysis code, and its prediction capability was validated by comparison with thoria-urania fuel irradiation test data. Analysis of the ThO2-UO2 fuel performance in a typical pressurized water reactor showed that it could be irradiated up to a burnup of 70 to 100 MWd/kg heavy metal by optimizing the fuel rod design, and the overall irradiation performance of the ThO2-UO2 fuel would be somewhat better than for UO2 fuel.

RAPID model to predict radial burnup distribution in LWR UO 2 fuel

The RAPID ( RA dial power and burnup P rediction by following fissile I sotope D istribution in the pellet) model was developed to predict the radial distribution of power, burnup and fissionable nuclide densities in the LWR UO 2 fuel pellets to be used in the fuel performance analysis code. It considers the specific radial variations of the neutron reactions of all the fissionable nuclides as functions of burnup and 235 U enrichment in the pellet. Comparison of the RAPID prediction with the measured data of the irradiated fuels and the predicted results by other codes showed good agreement, and therefore, the RAPID model can be used for UO 2 fuel of up to 10 w/o 235 U enrichment and 150 MWD/(kg U) pellet average burnup under the LWR environment.

Depletion Analysis of Mixed-Oxide Fuel Pins in Light Water Reactors and the Advanced Test Reactor

An experiment containing weapons-grade mixed-oxide (WG-MOX) fuel has been designed and is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The ability to accurately predict fuel pin performance is an essential requirement for the MOX fuel test assembly design. Detailed radial fission power and temperature profile effects and fission gas release in the fuel pin are a function of the fuel pin’s temperature, fission power, and fission product and actinide concentration profiles. In addition, the burnup-dependent profile analyses in irradiated fuel pins is important for fuel performance analysis to support the potential licensing of the MOX fuel made from WG-plutonium and depleted uranium for use in U.S. reactors.The MCNP Coupling With ORIGEN2 burnup calculation code (MCWO) can analyze the detailed burnup profiles of WG-MOX and reactor-grade mixed-oxide (RG-MOX) fuel pins. The validated code MCWO can provide the best-estimate neutronic characteristics of fuel burnup performance analysis. Applying this capability with a new minicell method allows calculation of detailed nuclide concentration and power distributions within the MOX pins as a function of burnup. This methodology was applied to MOX fuel in a commercial pressurized water reactor and in an experiment currently being irradiated in the ATR. The prediction of nuclide concentration profiles and power distributions in irradiated MOX pins via this new methodology can provide insights into MOX fuel performance.

MACSIS: A Metallic Fuel Performance Analysis Code for Simulating In-Reactor Behavior under Steady-State Conditions

Computational models for analyzing in-reactor behavior of metallic fuel pins in liquid-metal reactors under steady-state conditions are developed and implemented in the Metal fuel performance Analysis (computer) Code for Simulating the In-reactor behavior under Steady-state conditions (MACSIS). Sodium logging and constituent redistribution effects are considered in calculating the temperature profile. The model for the radial redistribution of the fuel constituent is based on the thermotransport theory. The fission gas release model takes multibubble size distribution into account to characterize the lenticular bubble shape and the saturation condition on the grain boundary. Finally, the clad strains are calculated from the amount of fission gas released and interface pressure. Sample calculations are performed to verify each model. The results show that in general, the predictions of MACSIS agree well with the available irradiation data.

Evaluation of the FRANCO finite element fuel rod analysis code

Knowledge of the temperature distribution in a nuclear fuel rod is required to predict the thermal and mechanical response of the strongly temperature-dependent fuel elements. In this research, the FRANCO ( F inite Element Fuel R od AN alysis CO de) computer code was developed for use on an IBM-PC to predict thermal and mechanical fuel performance characteristics for reactor operating conditions. A cross-sectional area of a fuel rod is discretized using constant strain triangular elements. To simplify the time-dependent fuel performance analysis, this code uses quasi-steady state conditions or slow power changes for time-independent problems; thus, time-dependent cases can be represented by quasi-static “snap shots.” The development and testing of the FRANCO program for deformation and heat transfer problems is described. The fuel performance models and solution procedures are provided in detail. The accuracy of FRANCO is examined through selected benchmark problems with other nuclear industry finite difference and finite element fuel analysis codes and actual Halden experimental measurements. These benchmarks show that FRANCO performs well except for the fuel-clad contact cases when compared to results from other fuel performance codes, such as FREY and ESCORE. The FRANCO program developed in this research effort is easy to run, fast, and yields comparable results.

Development of LWR Fuel Performance Analysis Codes

Abstract This paper is to cover brief history of the development of LWR fuel performance analysis codes in Japan. This includes the FEMAXI-III code which calculates local stress and strain of the cladding at power ramp due to the pellet cladding mechanical interaction (PCMI) using axisymmetric finite element method. From this code some improved codes were developed, such as FEMAXI IV, IRON and EIMUS. The FEMAXI-IV is developed by J AERI to add capacity of transient analysis and che IRON code was developed by Shikoku Electric Power Co. to analyze fuel reliability at load following operation and the EIMUS code was developed by CRIEPI for high burnup analysis. The major objective and contents of adapted submodels in these codes are presented with some calculational results. The future development and utilization of the codes, especially for performance analyses at high burnup, will be briefly discussed.

MIPAC, a computer code for fuel performance analysis by the finite element method

An axisymmetric finite element computer code named MIPAC has been developed for analysis of the mechanical interaction behaviour between a fuel pellet and cladding. This computer code can deal with elastoplasticity of the pellet and cladding materials, creep effects for the both materials, pellet-cladding and pellet-pellet contact problems, hot pressing effect of the fuel pellet, fuel pellet cracking, and the cracked pellet's stiffness. A cyclical boundary condition is introduced to deal with one pellet length instead of the full-size fuel rod. The contact problems are solved without a fictitious contact element. In the fuel pellet cracking model the crack opening and closing behaviour under arbitrary power changes can be treated by introducing five kinds of crack modes. Mismatch of irregular crack surfaces is taken into account in the evaluation of the cracked pellet's stiffness. Finally, calculated results are compared with experimental data to show validity of the computer code.

Mathematical description of WAFER-1, a three-dimensional code for LWR fuel performance analysis

Abstract This article describes in detail the mathematical formulation used in the WAFER-1 code, which is presently used for three-dimensional analysis of LWR fuel pin performance. The code aims at a prediction of the local stress-strain history in the cladding, especially with regard to the ridging phenomenon. To achieve this, a clad model based on shell theory has been developed. This model interacts with a detailed finite difference pellet model which treats radial and transversal cracking in the pellet in a deterministic way, based on certain assumptions with respect to the cracking pattern. Pellet and clad creep are taken into account. The inner core of the pellet, bounded by a specified isotherm, may be treated as a viscous material. Axial force exchange between pellet and clad is also included. The axial loading is distributed on the pellet end face with due regard to any pellet dishing. An arbitrary power history may be used as input to the model.