New irradiation facility for biomedical applications at the RA-3 reactor thermal column.

A new irradiation facility has been developed in the RA-3 reactor in order to perform trials for the treatment of liver metastases using boron neutron capture therapy (BNCT). RA-3 is a production research reactor that works continuously five days a week. It had a thermal column with a small cross section access tunnel that was not accessible during operation. The objective of the work was to perform the necessary modifications to obtain a facility for irradiating a portion of the human liver. This irradiation facility must be operated without disrupting the normal reactor schedule and requires a highly thermalized neutron spectrum, a thermal flux of around 10(10) n cm(-2)s(-1) that is as isotropic and uniform as possible, as well as on-line instrumentation. The main modifications consist of enlarging the access tunnel inside the thermal column to the suitable dimensions, reducing the gamma dose rate at the irradiation position, and constructing properly shielded entrance gates enabled by logical control to safely irradiate and withdraw samples with the reactor at full power. Activation foils and a neutron shielded graphite ionization chamber were used for a preliminary in-air characterization of the irradiation site. The constructed facility is very practical and easy to use. Operational authorization was obtained from radioprotection personnel after confirming radiation levels did not significantly increase after the modification. A highly thermalized and homogenous irradiation field was obtained. Measurements in the empty cavity showed a thermal flux near 10(10) n cm(-2)s(-1), a cadmium ratio of 4100 for gold foils and a gamma dose rate of approximately 5 Gy h(-1).

Preliminary liver dose estimation in the new facility for biomedical applications at the RA-3 reactor.

As a part of the project concerning the irradiation of a section of the human liver left lobe, a preliminary estimation of the expected dose was performed. To obtain proper input values for the calculation, neutron flux and gamma dose rate characterization were carried out using adequate portions of cow or pig liver covered with demineralized water simulating the preservation solution. Irradiations were done inside a container specially designed to fulfill temperature preservation of the organ and a reproducible irradiation position (which will be of importance for future planification purposes). Implantable rhodium based self-powered neutron detectors were developed to obtain neutron flux profiles both external and internal. Implantation of SPND was done along the central longitudinal axis of the samples, where lowest flux is expected. Gamma dose rate was obtained using a neutron shielded graphite ionization chamber moved along external surfaces of the samples. The internal neutron profile resulted uniform enough to allow for a single and static irradiation of the liver. For dose estimation, irradiation condition was set in order to obtain a maximum of 15 Gy-eq in healthy tissue. Additionally, literature reported boron concentrations of 47 ppm in tumor and 8 ppm in healthy tissue and a more conservative relationship (30/10 ppm) were used. To make a conservative estimation of the dose the following considerations were done: i). Minimum measured neutron flux inside the sample (approximately 5 x 10(9) n cm-2 s-1) was considered to calculate dose in tumor. (ii). Maximum measured neutron flux (considering both internal as external profiles) was used to calculate dose in healthy tissue (approximately 8.7 x 10(9) n cm-2 s-1). (iii). Maximum measured gamma dose rate (approximately 13.5 Gy h-1) was considered for both tumor and healthy tissue. Tumor tissue dose was approximately 69 Gy-eq for 47 ppm of (10)B and approximately 42 Gy-eq for 30 ppm, for a maximum dose of 15 Gy-eq in healthy tissue. As can be seen from these results, even for the most conservative case, minimum tumor dose will be acceptable from the treatment point of view, which shows that the irradiation conditions at this facility have quite good characteristics for the proposed irradiation.

Monte Carlo calculation for the development of a BNCT neutron source (1eV – 10 KeV) using MCNP code

Different materials have been studies in order to produce the epithermal neutron beam between 1eV - 10 KeV, which extensively used to irradiate patients with brain tumors such as (GBM). For this purpose we have studied three different neutrons moderators (H2O, D2O and BeO) and their combinations, four reflectors (Al2O3, C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation show, That the best obtained assembly configuration correspond to the combination of the three moderators H2O, BeO and D2O jointly to Al2O3 reflector and two filter Cd +Bi optimize the spectrum of the epithermal neutron at 72 percent, and minimize the thermal neutron to 4 percent and thus it can be used to treat the deep brain tumor. Our calculations have been performed by means of the Monte Carlo N- particle code MCNP 5C.

Comparison of the performance of two NCT treatment planning systems using the therapeutic beam of the RA-6 reactor.

This work evaluates the performance of two NCT treatment planning systems: NCTPlan, developed by the CNEA and the Harvard-MIT group, and SERA, developed by the INEEL/Montana State University group. The study was performed in some simple geometries with the therapeutical hyperthermal beam of the RA-6 facility at Bariloche, Argentina. The first geometry was a rectangular phantom and calculations and measurements were made along the central beam axis and along a parellel axis, 4 cm apart from the central beam axis. Measurements and calculations were also performed in a cylindrical phantom, to explore the behavior of the treatment planning systems in a geometry simulating an extremity, in accordance with the CNEA clinical protocol. Comments on differences in source definitions and cross sections libraries are also included in the text. It can be seen that both codes give acceptable results on the central beam axis and on a lateral axis, showing good agreement with experimental results.

Implantable self-powered detector for on-line determination of neutron flux in patients during NCT treatment.

A novel system to determine thermal neutron flux in real time during NCT treatments was developed in the National Atomic Energy Commission of Argentina. The system is based on a special self-powered detector that can be implanted in patients owing to its small size and biocompatibility. High voltage is not required to operate this kind of detectors, which is a considerable advantage in terms of medical uses. By choosing the appropriate materials, it was possible to obtain a prototype with thermal neutron sensitivity providing for an adequate signal level in typical NCT thermal fluxes. It was also possible to minimize gamma response in order to neglect its contribution.