Multifunctional composites made with boron as the absorbers of low energy neutrons could be used for structural shielding materials. These composites could potentiallyproduced having the flame retardancy properties. Many investigators have developed polymer composites specificallyfor their radiation shielding properties. Boron carbide is amaterial that is known for its hardness, third only to diamond and cubic boron nitride. Increased mechanical strength, high melting point, and low specific gravity are also properties that make it an attractive material [1]. Boron has a high thermalneutron absorption cross section of 767 barn and lithium and helium as its decay products. This decay products are desirable because of their short half-lives and non-radioactivecharacteristics [2]. However, boron is brittle and difficult to produce in shapes. Boron in the form of boron carbide makes it as a betteralternative. Polyethylene is also a well known material for shielding applications because of its high hydrogen content [3]. A composite combining the shielding characteristics of boron carbide and polyethylene could provide superior shielding properties than neat polyethylene. In thedevelopment of composite materials, the reinforcements currently used range decrease in size from macro-, over mesoandmicro- to the nanoscopic scale. The identified mechanisms and the conventional theories explain the observed influence of micro-sized reinforcement and itsdispersion in the matrices, mechanical and physical properties of composites, may not be valid for nano-fillers anymore [4]. The interest of scientists in applying nano-scaled fillers into polymer matrices is the attainment of potentially unique properties, as a result of nanometric dimensions. The main objective of this paper is to investigated upon the effects of particle size and weight percent of the reinforcement phase onthe absorption ability of thermal neutron by HDPE/B4C composites as radiation shielding. The maintained goal was studied By means of monte carlo simulation method utilizing the MCNP1 computer code. The MCNP code, an internationally recognized code for analyzing the transport ofneutrons and gamma-rays using the Monte-Carlo method, is developed and maintained by Los Alamos National Laboratory [5].