Radiation protection basis of design for the South African isotope facility - SAIF

To expand on its research agenda and meet the growing demand for nuclear medicine products, the South African Isotope Facility (SAIF) was launched. This study focuses on Radiation Protection aspects for SAIF Phase 1, specifically the verification of existing- and basis of design for supplementary engineering controls for the retrofitted accelerator complex, housing an IBA C70 Cyclone 70 MeV H-minus cyclotron and four in-house designed high intensity particle beam target stations. For the dosimetric assessment, a selection of calibrated instruments were used to determine the ambient dose equivalent rate H*(10) at points of interest. Physical dose rate measurements were performed for various source terms (66 MeV p+ on Cu, Al, Mg, Ga and Li) and beam intensities up to tens of micro Amperes, beyond in-situ bulk shielding material, for compound shielded target stations and streaming through existing labyrinths. Comparisons are drawn between actual measurements and the FLUKA code. Primarily to determine the in-situ shielding characteristics of unverified bulk shielding material cast in the 1980s, as core drill samples yielded vastly divergent aggregate types and material densities.


Introduction
The NRF iThemba LABS embarked on a strategy to expand the research agenda of the facility and simultaneously exploit the growing global demand for nuclear medicine products through the development of the South African Isotope Facility (SAIF).A 70 MeV H-minus cyclotron, producing intense proton beams at currents up to 750 µA -dual extraction channels capable of delivering 375 µA proton beams concurrently -has been acquired to offset radionuclide production activities from the K200 separated sector cyclotron (SSC).The development is to be integrated within the existing building infrastructure to radically increase particle beam availability for research activities and ramp up radionuclide production output [1].
Three decommissioned hadron therapy bunkers located in close proximity to the present radioisotope production infrastructure were identified as the new home for SAIF.Implementation of a development of this nature and scale in repurposed bunkers, can potentially achieve a significant cost saving on building infrastructure, as bulk shielding material costs are prohibitively high.The downside is the complexity of retrofitting the accelerator-, beam transport-and target station mega structure into an existing building where structural modifications are severely constrained -a trade-off between useable vault space and shielding elements, as dictated by radiation protection requirements.
As part of this study, shielding design elements during the various design phases were verified using Monte Carlo type radiation transport simulations.The calculations were in turn benchmarked against

Method 2.1 Source terms
The following source terms were generated A) 66 MeV p+ on Cu (thick target), with no local shielding, B) 66 MeV p+ on Al, Mg, Ga (thick compound target), housed within the 28-ton horizontal beam target station (HBTS) consisting of concentric layers of steel, borated wax and lead, and C) 66 MeV p+ on Li (thin target), with no local shielding in place.

Physical dose equivalent rate measurements
Two identical NM2 neutron monitors (Nuclear Enterprises Ltd., Edinburgh, UK) were employed to measure neutron ambient dose equivalent rates Ḣ*(10) for the various source terms.Measurements were conducted for source terms A & B at each diagonal intersection as well as entrance and exit point, along a 4-legged vault access labyrinth.For source term C, measurements were conducted at 90º for varying beam intensities of several micro Amperes beyond in-situ bulk shielding material.

Radiation transport simulations
The radiation transport code FLUKA was used to simulate Ḣ*(10) at various points of interest for the defined source terms [2].To speed up calculations in bulk shielding material for this extensive geometry, region importance BIASING was applied [3].This was supplemented by performing a two-step secondary source sampling operation for the HBTS.The EWT74 option which applies the WORST exposure geometry was selected to derive Ḣ*(10) [4].

Results and Discussion
Isodose-and 2-D plots representative of the FLUKA simulated neutron ambient dose equivalent profile were generated for the various source terms and measurement geometries.Physical dose rate measurements for fixed locations, determined through superimposition of simulated geometry onto facility layout line drawings, served as standard for comparisons.

Labyrinth streaming
FLUKA simulated neutron Ḣ*(10) for source term A) p(66)/Cu, evaluated at various locations along the labyrinth are in agreement (P-value = 0.008) with direct measurements, albeit systematically overestimated see Figure 1.Similarly, Ḣ*(10) observed for source term B) p(66)/compound target in HBTS are not significantly different (P-value = 0.041) when compared to FLUKA values for the same target and geometry Figure 1.
A systematic reduction in the neutron Ḣ*(10) streaming through the labyrinth in the order of 1x10 -3 was both calculated and observed for the locally shielded source, compared to the same beam characteristics for p(66)/Cu in an unshielded geometry.This agrees with the optimized shielding design of the HBTS for 70 MeV proton beams [5].
A peak value of 2.31 ± 0.05 µSv/h was recorded at 90º beyond 1.5m heavily reinforced concrete beams.Correspondingly the calculated dose rate max of 5.51 ± 0.46 µSv/h beyond the shielding was found to be located at <90º.A forward biased spectrum is typically associated with thin targets.For this simple geometry, the shielding efficacy is under estimated by a factor greater than 2.

Conclusion
FLUKA simulated values across all but one measurement location and for various beam target combinations, agree to within an order of magnitude and importantly underestimates the in-situ shielding efficacy, providing a conservative tool for shielding design in this proton energy range.Calculations for compound targets, locally compound shielded, transported through a vast space and an extensive labyrinth, has demonstrated the functionality and reliability of BIASING and two-step simulations, for medical isotope production facility shielding design.

Figure 1 .
Figure 1.Neutron dose equivalent rates, both calculated and measured at various locations along the horizontal beam target station vault access labyrinth for p(66)/HBTS compound target with local compound shielding and p(66)/Cu with no local shielding in place.Measurements were conducted using two identical NM2 neutron detectors and the Monte Carlo type simulation package FLUKA, to calculate neutron dose equivalent rates.

Table 1 .
Direct neutron dose equivalent rates measured beyond 1.5m concrete shielding at 90º for an incrementally rising beam current p(66)/Li (thin target).