Takanori Kameyama

Concerning the microstructure changes that occur at the surface of UO2 pellets on irradiation to high burnup

Abstract It is shown that in addition to the precipitation of small gas filled pores, there is a pronounced reduction in the grain size at the surface of UO2 fuel at high burnup. These microstructure changes were first observed when the local burnup exceeded 70–80 MWd/kgM. Generally, the change in microstructure does not penetrate more than 200 μm. However, in a HWR fuel irradiated to 75 MWd/kgM a large part of the pellet cross section was found to have been affected. Temperature predictions for this fuel suggests that it is the restructuring accompanying thermally activated fission gas release at 1100 to 1200°C that limited the distance over which the microstructure changes occur. Apparently, the formation and fission of Pu is not directly responsible for the change in fuel microstructure. The porosity evidently contains part of the fission gas that is lost from the UO2 lattice in the region where the microstructure changes take place.

Proposal of Direct Calculation of Kinetic Parameters βeff and Based on Continuous Energy Monte Carlo Method

Direct calculation methods of kinetic parameters are proposed based on the continuous energy Monte Carlo method. In the proposed methods, the effective delayed neutron fraction βeff and the neutron generation time ∧ are estimated using eigenvalue calculations. The expected number of fission neutrons in the next generation is newly applied to the proposed methods instead of the adjoint flux that has been conventionally used. The algorithms to estimate the kinetic parameters are established and incorporated into the continuous energy Monte Carlo transport calculation code MCNP-4C, which is versatile for eigenvalue calculations of nuclear reactor cores with various types of neutron energy spectrum and geometry. The proposed methods were validated since the calculated values agreed with the experimental data of βeff and βeff/∧ for the critical cores within accuracies of 4.5% and 10%, respectively.

Development of Calculation Technique for Iterated Fission Probability and Reactor Kinetic Parameters Using Continuous-Energy Monte Carlo Method

New algorithms and techniques are developed for calculating the iterated fission probability (IFP ) using a Monte Carlo eigenvalue calculation scheme. The proportionality of the calculated IFP to the adjoint flux is confirmed numerically. This IFP is then used with the MCNP code and its point-wise cross section data libraries to calculate kinetic parameters weighted by the adjoint flux. Experimental data for critical cores validate these theoretical estimates.

Temperature and fission rate effects on the rim structure formation in a UO2 fuel with a burnup of 7.9% FIMA

Abstract A BWR design UO2 fuel irradiated to a burnup of 7.9% FIMA was selected for a careful calculational and experimental analysis because the rod experienced an unusual power history: it had two high power periods at 1.7% FIMA and between 4 and 5% FIMA causing increased fuel temperatures and thus increased gas release and damage recovery. As a consequence, two parameters generally considered to be important for grain subdivision (rim structure formation) were locally different from normal fuel, i.e. fission gas inventory and extent of radiation damage. Histories of temperature, fission rate and fission gas release were calculated at different radial positions. Microstructure observations (TEM, SEM) revealed the typical high burnup grain subdivision process (polygonization) which extended to a maximum of 1.65 mm (r/r0 = 0.73) from the pellet surface inwards. For this radial position, the calculations yielded a local temperature of 1200°C and predicted that more than half of the fission gas was released during the second high power period for this radial position. The results give thus information on the importance of the fission gas inventory for the burnup threshold of restructuring.

Numerical analysis for microstructure change of a light water reactor fuel pellet at high burnup

The peripheral region of a high burnup light water reactor (LWR) fuel pellet shows a microstructure that is different from the as-fabricated microstructure. The region where the microstructure change occurs (the rim region) is highly porous, and the original grains in the rim region are divided into much smaller subgrains. The electron probe microanalysis data of high burnup fuels indicate fission gas depletion in the rim region as well as in the central region. The burnup in the rim region is enhanced by built-up plutonium derived from a 238 U self-shielding effect, which is called a rim effect. The rim effect accelerates microstructure change in the peripheral region. We developed a detailed burnup analysis code ANRB computing the rim effect in LWR fuels

The FLEXBURN neutron transport code developed by the Sn method with transmission probabilities in arbitrary square meshes for light water reactor fuel assemblies

The FLEXBURN neutron transport code is developed by the discrete ordinates (S{sub n}) method to analyze heterogeneous fuel assemblies in light water reactors. The transport equations are formulated with transmission and leakage probabilities in arbitrary convex square meshes. Arbitrary convex square meshes precisely describe fuel assemblies as lattices of cells. The code deals with fuel assemblies including gadolinia doped fuel rods, water rods, or plutonium mixed fuel rods with control blades. The code can make burnup calculation sequentially to high burnup. The results computed by the FLEXBURN code are validated by comparing them with those of the ANISN typical transport code and the KENO-IV Monte Carlo code. The FLEXBURN code provides control blade worth and detailed distributions of flux, power, burnup, and atomic densities in complicated boiling water reactor and pressurized water reactor fuel assemblies.