The Malaysian 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the calculation of neutron flux and power distribution in PUSPATI TRIGA REACTOR (RTP) 14th core configuration. The 3-D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA core and fuels. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data as well as S (α, β) thermal neutron scattering functions distributed with the MCNP code were used. Results of calculations are analyzed and discussed.
The Malaysian’s PUSPATI TRIGA Reactor (RTP) achieved its initial criticality on June 28, 1982. The reactor is designed to effectively implement various fields of basic nuclear research, manpower training, and production of radioisotopes. Several past activities on neutronics modelling development and validation of the RTP were carried out using Monte Carlo Code MCNP. In this work, the developed model was used to characterise in-core and beam-ports irradiation facilities of the reactor. The thermal and fast neutron flux distributions in these facilities were determined using MCNP mesh tally method. It was found that the flux as well as its spectral characteristics depended very much on the position of the irradiation facility in the reactor core or in the beam-ports. The maximum neutron flux was found to be in the Central Thimble facility with 1.98E13 nv of thermal neutron. The thermal-to-total flux ratio varies significantly from 0.41 for the in-core facility, 0.58 in the reflector and up to 0.88 in the beam-ports.
The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial
criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of
basic nuclear research, manpower training, and production of radioisotopes. This
paperdescribes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP);
focusing on the application of the developed reactor 3D model for criticality calculation,
analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D
continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full
model of the TRIGA reactor. The consistency and accuracy of the developed RTP MCNP model
was established by comparing calculations to the experimental results and TRIGLAV
code.MCNP and TRIGLAV criticality prediction of the critical core loading are in a very good
agreement with the experimental results.Power peaking factor calculated with TRIGLAV are
systematically higher than the MCNP but the trends are the same.Depletion calculation by both
codes show differences especially at high burnup.The results are conservative and can be
applied to show the reliability of MCNP code and the model both for design and verification of
the reactor core, and future calculation of its neutronic parameters.
In order to prepare Malaysia to be nuclear ready, the Malaysian 1 MW TRIGA MARK II research
reactor (RTP) located at the Malaysian Nuclear Agency was premeditated with the aim to effectually
actualize the multitude areas of basic nuclear research, labor training and education. To meet the
modern safety standards, analyses of a strong interaction between the thermal-hydraulic system
behavior and the space-dependent neutron kinetics are needed as mere thermal-hydraulics codes are
said to be incapable to succeed the present safety standards. This could be achieved through the
coupling of neutronic and thermal-hydraulic codes of the reactor. Previous studies had shown that the
coupled codes are able to successfully be employed for the correlation between thermal-hydraulic
analysis and neutron kinetics at transient and steady state. In this study, the coupling was achieved
through MCNP and TRIGLAV codes for neutronic and thermal-hydraulic respectively. Core-15 of
RTP was modeled for both of the codes; hence calculating the criticality, analysis of power and
neutron flux distribution. The consistency and accuracy of the developed Core-15 MCNP model was
established by comparing calculations to the experimental results and TRIGLAV code. The criticality
predictions for both codes are in very good agreement with the experimental results. The core reached
its criticality after 66 fuels. The highest hot rod power peaking factor was found to be 1.28. The
results are conservative and can be applied to show the reliability of MCNP and TRIGLAV codes.
The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II REACTOR were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06E8 n/cm2/s. According to IAEA (2001) flux of 1.00E9 n/cm2/s requires three hours of treatment. Few modifications were needed to get higher flux.
Thermal neutron beam from thermal column was selected for a Boron Neutron Capture Therapy
(BNCT) system utilizing the Malaysian TRIGA MARK II reactor. Determination of shielding
materials for fast and epithermal neutron was conducted. The materials selected were polyethylene,
paraffin and water. For gamma-ray shielding, lead was used. The objective of this paper is to present
the simulation and verification of an optimal design of BNCT collimation at a beam. line viewing the
thermal column. A collimator was made from polyethylene pipe with 8 cm of diameter filled with
paraffin.