THE USE OF MONTE CARLO CODE FOR RADIATION TRANSPORT AND DOSIMETRY CALCULATION FOR NEUTRON RADIOGRAPHY EXPOSURE ROOM FACILITY AT REACTOR TRIGA MARK II PUSPATI (RTP) 1,*Rafhayudi Jamro, 1 Muhammad Rawi M. Zain, 2Redzuwan Yahaya, 3Abdul Aziz Mohamed, 1Megat Harun Al Rashid Megat Ahmad, and 1Hafizal Yazid, 1Agensi Nuklear Malaysia, 43000 Kajang, Selangor, Malaysia 2 Nuclear Science Program, Universiti Kebangsaaan Malaysia, 43000 Bangi, Selangor, Malaysia, 3 Uniten, Kajang Selangor, *Contact person: rafhayudi@nuclearmalaysia.gov.my RESULTS AND DISCUSSION The input of MCNP5 included 5010 cycles made of 10 inactive cycles and 5000 active cycles with 10,000 histories per cycle using KCODE card. The flux and dose distribution in the NR collimator is shown Fig.3 and Fig.4 respectively. Fig 3: Flux distribution inside NR Collimator Fig.4: Dose distribution inside a NR In this project, neutron flux and dose at neutron radiography (NR) facility of TRIGA Mark II PUSPATI (RTP) was modeled and simulated using MC approach. The Monte Carlo simulation of the NR was presented. The Monte Carlo technique gave reasonable picture of flux and dose distributions inside the NR collimator. 1SAR Report For Puspati Triga Mark II Reactor Facility, Pusat Penyelidikan Atom Tun Dr Ismail (Puspati), Bangi, Selangor, 1983 2MCNP-A general Monte Carlo N-Particle Transport Code, Version LA-UR-03-1987. X-5 Monte Carlo Team. Los Alamos National Laboratory. 2003 INTRODUCTION Interactions of nuclear particles with matter can be described through statistical means. The stochastical modeling of this nuclear interactions can be best simulated using Monte Carlo approach. This Monte Carlo (MC) method is a computational algorithm that can provide approximate solutions to a variety of nuclear and also other physical problems by the simulation of random quantities. A Monte Carlo simulation of photon and neutron flux at the neutron radiography exposure room (NuR II) with various shielding material inside a NR collimator at Malaysian Nuclear Agency was performed using the MCNP5 code. The objective of this work is to model the NuR II collimator in conjunction with for obtaining radiation transport and dosimetry calculation result. NR DISCRIPTION NR is a well know NDT used to detect the presence and structural nature of materials opaque to neutron. The overview of a neutron collimator of NUR II is shown in Fig.1. Step divergent type of collimator is used with its length 220 cm with 10 cm and 20 cm in diameter respectively. Thus, the collimation ratio is 75 and thermal neutron flux is around 1.04x105ncm-2s-1 [1]. NR collimator was cladded with stainless steel. There are a Bismuth, lead and flexi boron inside the NR collimator for shielding purposed. METHODOLOGY Monte Carlo neutron transport simulations was carried out using MCNP5 computer program. Neutron cross section of the materials were taken from the continuous-energy and discrete neutron data libraries (ENDF/B VI.6 libraries) [2]. The user creates an input file that is subsequently read by FOTRAN 90 compiler in MCNP. The MCNP5 code was run using stand alone 3.00GHz CPU with 2 GB RAM under Microsoft operating system. The MCNP simulation was run with F4 tally. The F4 tally was also modified by using the dose function card and flux to dose conversion factor. The MCNP5 is employed to construct a complete model of NR collimator positioned accurately inside a reactor building. NR Collimator was represented as cylinder of appropriate materials and dimension, positioned as shown in fig. 2. Fig. 1: Overview of the NuR II collimator at Malaysian Nuclear Agency Fig.2: Radial model for the NR Collimator in MCNP5