The Application of the Combined Fission Matrix Theory in Fast Reactors

The combined fission matrix theory is a recently-developed hybrid neutron transport method. It features high efficiency, fidelity, and resolution whole-core transport calculation. The theory is based on the assumption that the fission matrix element ai,j is dominated by the property of the destination cell i. This assumption can be well explained in thermal reactors, and the combined fission matrix method has been validated in a series of thermal neutron system benchmarks. This work examines the feasibility of the combined fission matrix theory in fast reactors. The European Sodium Fast Reactor is used as the numerical benchmark. Compared to the Monte Carlo method, the combined fission matrix theory reports a 64 pcm keff difference and 8.3% 2D RMS error. The error is much larger than that in thermal reactors, and the correction ratio cannot significantly reduce the material discontinuity error in fast reactors. Overall, the combined fission matrix theory is more suited for thermal reactor transport calculations. Its application in fast reactors needs further developments.

Assessment of BSL kω Turbulence Model for Predicting Flow Characteristics in a Solitary Wave Motion

A deep understanding of boundary layer characteristics under wave motion, especially bottom shear stress on the seabed, is critical in sediment transport calculation and modeling. In this present study, boundary layer characteristics under solitary wave motion over the smooth bed are investigated through the Baseline (BSL) kω model. It will also consider the result of a laboratory experiment performed in a closed conduit generation system. The conclusion said that the BSL kω model can replicate well on horizontal and vertical velocity distribution at Reynolds number (Re) = 2.25 x 105. In addition, a turbulent spike which was occurred during decelerating phases of oscillatory motion can be predicted well also by the BSL kω model at Reynolds number (Re) = 6.06 x 105.

The Use of the MCNP Code for Radiation Damage Calculations

This work gives a detailed analysis of the result of Monte Carlo physics practical using MCNP. This paper describes basic concepts of the Monte Carlo theory of radiation transport calculation and also discusses the variance and the history method as used in Monte Carlo Problem solving. Therefore, in this exercise the MCNP code has been used to solve and estimate the number of neutron flux. The paper investigated the impact of the primary radiation damage in iron by the neutron energy irradiation. The established measurement of radiation damage is the displacements per atom (dpa) in matter as a function of neutron energy. The simulations were carried out to calculate the dpa cross section.