Monte Carlo calculation method can be used for resolving particle transport in matter, and particularly the transport of neutrons in the environment of the reactor core. The method has become more efficient because of high accuracy of updated nuclear data and fast development of advanced super-computing system. In this work, we would like to present calculations for kinematic characteristics of neutron transport in a typical configuration of the pressurized water reactor (PWR) fuel assembly based on the Monte-Carlo simulation method. We concentrate in two main results: (1) neutron energy spectrum at fuel rod and (2) optimal thickness of light water reflector. | Communications in Physics, Vol. 25, No. 1 (2015), pp. 91-96 DOI: SIMULATION FOR NEUTRON TRANSPORT IN PWR REACTOR MODERATOR AND EVALUATION FOR PROPER THICKNESS OF LIGHT WATER REFLECTOR NGUYEN TUAN KHAI AND PHAN QUOC VUONG Institute for Nuclear Science and Technology, 179 Hoang Quoc Viet, Hanoi E-mail: ntkhai@ Received 26 November 2014 Accepted for publication 12 December 2014 Abstract. Monte Carlo calculation method can be used for resolving particle transport in matter, and particularly the transport of neutrons in the environment of the reactor core. The method has become more efficient because of high accuracy of updated nuclear data and fast development of advanced super-computing system. In this work, we would like to present calculations for kinematic characteristics of neutron transport in a typical configuration of the pressurized water reactor (PWR) fuel assembly based on the Monte-Carlo simulation method. We concentrate in two main results: (1) neutron energy spectrum at fuel rod and (2) optimal thickness of light water reflector. Keywords: Monte-Carlo method, neutron transport, pressurized water reactor (PWR), fuel rod, fuel assembly, light water reflector. I. INTRODUCTION Fission reaction takes place when a heavy nucleus as uranium, thorium and plutonium absorbs neutron, high energy gamma radiation or even charged particles and is splited into two intermediate-mass fragments with releasing a few new neutrons under prompt and delayed types. As known, amongst of these the fission of the heavy actinide isotopes as 235 U,239 U and 241 Pu induced by thermal neutrons is the most important process to generate power in nuclear reactor. It is obvious that in nuclear reactor the neutron generation from fission and their transport including slowing down in moderator and absorption in fuel are very essential as basis of reactor physics and kinetics. The basis of transport theory can be derived from the particle density .
