Neutron Field Modification with Various Materials

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Introduction

Neutrons have inherent properties that allow them to have wide-ranging applications in various disciplines. Currently, the major neutron sources in the Philippine setting are medical accelerators where neutrons are produced as a byproduct when accelerated charged particles with sufficient energy hit heavy targets. Neutron field characterization is essential to evaluate the potential neutron dose to radiation workers and the public for these applications. However, it is only recently that neutron field characterization was performed in local medical accelerator facilities. To date, only one PET cyclotron facility and one linear accelerator facility has neutron field data. But neutron fields tend to vary depending on facility design, which includes the facility dimensions, the materials available in the facility, as well as the composition of materials in the facility. Each medical accelerator facility that is operating under certain conditions will have a unique neutron field, thus warranting the importance of neutron field characterization, at least once, for medical accelerator facility with neutron byproducts. In this work, we simulate the effect of different materials on the neutron field that is produced by a ²⁵²Cf source in the PNRI Neutron Laboratory (PNL) to demonstrate the significance of neutron field characterization for facilities with varying configurations.

Materials and Methods

A detailed MCNP model of the ²⁵²Cf source in the PNL irradiation room is prepared based on manufacturer specification and facility drawings. Material compositions in the model are based on the PNNL compendium. A 10 cm x 20 cm x 20 cm cell that is positioned at the same level and at 5 cm distance from the source was added in the computational model. The material declared in this cell is varied to investigate the effect of different materials on the PNL neutron field. The materials considered for this preliminary investigation are aluminum, concrete, and polyethylene (PE), which are commonly found in medical accelerator facilities. The FMESH tally of MCNP is used to calculate the horizontal and vertical neutron distribution in a 300 cm x 300 cm x 20 cm meshes that are centered at 50 cm from the source. The simulated neutron fields obtained from different materials are then compared with the neutron field when the cell material is declared as air.

Results and Discussion

Figure 1 (a) and (c) show the simulated neutron field when the material declared in the cell is air, while Figure 1 (b) and (d) results for PE. The field "shadow" cast by the cell in the PE-modified field demonstrates that a small volume of this material results in significant change in the field. The resulting field was also changed with the introduction of aluminum and concrete.

Conclusions

This work demonstrates that neutron fields can be modified significantly with the presence of small volume materials that are found in medical accelerator facilities. The result of this work will be used as a basis for experiments that will be performed at the PNL to investigate the limitations of commercial neutron survey meters when conducting dose measurements in different neutron fields.