CALCULATION OF NEUTRON FLUX DISTRIBUTION AT PIERCING BEAM PORTS OF PLATE TYPE RESEARCH REACTOR BANDUNG. Based on a strategic plan of TRIGA 2000 Bandung's future operation, BATAN has already decided to implement an option to convert the fuel elements core of TRIGA 2000 from using the cylindrical type of elements produced by General Atomic to MTR plate type of fuel elements produced by local fuel element manufacture. The core design calculation has proved that the core configurations of 5 x 5 matrix using local plate type fuel elements met the requirement of core neutronics design. In addition to the current core configuration, further study must be added to consider the use of beam ports as utilization facilities in the design. The neutron flux distribution at piercing beam port has been calculated based monte carlo algorithm using TRIGA MCNP and MCNP software. The calculation result showed that at piercing beam port surface neutron flux distribution is not quite symmetric. The highest neutron flux at piercing beam port is 9.4 × 10 8 (2. ⁄), where as the flux of neutron thermal energy group is 3.54 × 10 8 (2. ⁄). These results are considerably appropriate for such core configuration and as a result, they can be used as a basic data for designing Plate Type Research Reactor Bandung, especially for neutron diffraction experiment

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Design of Irradiation Facilities at Grid E-1 of Plate Type Research Reactor Bandung. Plate Type Research Reactor Bandung (PTRRB) core design is one of the result of PTRRB research programs. In the previous study the irradiation facilities at grid E-1 has not been designed and also distribution of thermal, epithermal and fast neutron flux at grid E-1 has not been studied. Since that data is very important especially in radioisotope production and neutron beam tube analysis, therefore in this study irradiation facilities at grid E-1will be designed. Previous PTRRB core design is a base for designing irradiation facilities at grid E-1. Considering geometrical of grid E-1 and aluminum tube dimension there are three possibilities aluminum tube configuration. The configurations are configuration 1, 2 and 3. Each configuration was modelled as arrangement of four aluminum tubes and each tube filled by four aluminum irradiation capsules. That configuration was starting point to made MCNP PTRRB reactor core model so there are three MCNP PTRRB reactor core model. MCNP PTRRB reactor core model is needed because MCNP software are computer program for calculating excess reactivity and neutron flux distribution at grid E-1. Result excess reactivity calculation of three configuration indicate that after installing irradiation tube excess reactivity is lower than of limit excess reactivity value 10.9 % of neutronic safety criteria of PTRRB design. Based on neutronic safety criteria, the three configuration is accepted for irradiation facilities PTRRB. Neutron flux calculation result of three configuration reveals that the highest neutron flux is located at capsule no II and III. Profile of thermal neutron flux, epithermal neutron flux and fast neutron flux of three configurations are similar. Neutron flux of thermal, epithermal and fast neutron of three configuration are slightly different. The calculation result reveal that highest thermal neutron flux at grid E-1 is 2.70 × 1013(n/cm 2.sec) at configuration 2. Based on neutronic safety criteria and thermal neutron flux, configuration 2 is appropriate for irradiation facilities of PTRRB.

One of the results from Plate Type Research Reactor Bandung (PTRRB) research program is PTRRB core design. Previous study on PTRRB has not calculated neutron flux distribution at its central irradiation position (CIP). Distribution of neutron flux at CIP is of high importance especially in radioisotope production. In this study, CIP was modeled as a stack of four to five aluminum tubes (AT), each filled by four aluminum irradiation capsules (AIC). Considering AIC dimension and geometry, there are three possibilities of AT configuration. For irradiation sample, 1.45 gr of molybdenum (Mo) was put into AIC. Neutron flux distribution at Mo sample was calculated using TRIGA MCNP and MCNP software. The calculation was simulated at condition when fresh fuel is loaded into reactor core. Analyses of excess reactivity show that, after installing irradiation AT and Mo sample was put into each configuration, the excess reactivity is less than 10.9 %. The highest calculated thermal neutron flux at Mo sample is 5.08×1013 n/cm2.s at configuration 1. Meanwhile, the highest total neutron flux at Mo sample is located at capsule no. II and III. Thermal neutron flux profile is the same for all configurations. This result will be used as a basic data for PTRRB utilization.Keywords: Central Irradiation Position, Neutron Flux Distribution, MCNP, PTRRB