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Salehi et al., 1:11http://dx.doi.org/10.4172/scientificreports.537

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Volume 1 • Issue 11 • 2012

Keywords: BNCT; Neutron filter; Epithermal neutron; Fluental

IntroductionBoron Neutron Capture Therapy (BNCT) is an experimental binary

cancer radiotherapy modality in which a boronated pharmaceutical that preferentially accumulates in malignant tissue is first administered, followed by exposing the tissue in the treatment volume to a thermal neutron field. BNCT of melanoma depends on the selective loading of the tumor with the enriched 10B Compound and subsequent irradiation with thermal neutron.

The high thermal cross section for the 10B(n,α)7 Li reaction means a high production rate of α and 7Li particle. These charged particle have range in tissue of <10 μm, hence most of the energy generated in the tumor cell is absorbed within that cell.

For the BNCT of deep seated tumors, an epithermal neutron beam is considered to be superior to a thermal beam because of its deeper penetration and skin sparing properties. Such a beam would be thermalised within tissue itself. Boronated cells are selectively destroyed via energy deposition resulting from the 10B(n,α)7 Li interaction [1].

A conceptual design to produce epithermal photoneutrons by high energy photons (due to bremsstrahlung) impinging on deuterium targets is presented along with computational and experimental neutron production data.

A clinically acceptable, filtered epithermal neutron flux of the order of 107 neutrons per Second per milliampere of electron current is shown to be obtainable. The useful neutron energy range for BNCT purposes is from 1 eV to 10 keV (sometimes 40 keV is as-summed).

The effectiveness of the treatment is expected to increase both the amount of boron in the cells and the neutron flux increase. In this treatment, advantage is taken of the nuclear reaction between boron and thermal neutrons, resulting in cell death from the energy deposited along the paths of resultant alpha particles and lithium ions.

The main tasks are to deliver enough boron and thermalized neutrons to each tumor cell to ensure cell death. In order to realize the benefits of NCT (Neutron Capture Therapy), a sufficiently intense and pure epithermal neutron source is critical.

*Corresponding author: Danial Salehi, Faculty of Engineering, Science and Research Branch, Islamic Azad University, P.O. Box 14515-775, Tehran, Iran, E-mail: da.salehi@yahoo.com

Received June 27, 2012; Published September 28, 2012

Citation: Salehi D, Sardari D, Salehi M (2012) Evaluation of Design Neutron Filters in BNCT. 1:537 doi:10.4172/scientificreports.537

Copyright: © 2012 Salehi D, et al. This is an open-access article distributed under the terms of the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original author and source are credited.

AbstractThis paper reviews the development of neutron filters using Boron Neutron Capture Therapy (BNCT). At present

BNCT research necessity of an epithermal beam is to generate the necessary thermal neutron field at the desired depth. Through the use of an epithermal beam, deeper-seated tumors can be treated more effectively [1]. Epithermal neutron beams can be generated by small nuclear reactors and by accelerator based neutron sources. So far, only reactors have been actually used to produce therapeutically useful epithermal neutron beams for BNCT; accelerator-based epithermal neutron sources may constitute the basis for a more deployable technology for BNCT than reactor-based sources in the long term [2]. Accelerators offer a number of potential advantages over reactor-based neutron sources for clinical applications [3]. The goal of reactor epithermal neutron filter conceptual design effort was to provide a low cost, simple filter design that is easy to fabricate and install while meeting or exceeding the nuclear performance requirements of the BNCT program. Therefore, in this work the use of aluminum compound, teflon compound and fluental and heavy water was considered.

Evaluation of Design Neutron Filters in BNCTDanial Salehi1*, Dariush Sardari1 and Milad Salehi2

1Science and Research Branch, Islamic Azad University, P.O. Box 14515-775, Tehran, Iran2Boroujerd Branch, Islamic Azad University, Boroujerd, Iran

Material and MethodNeutrons from the neutron source have energies in the fast energy

range (average energy of around 2 MeV), and these neutrons will give a toxic dose to the human body; therefore the fast neutrons must be slowed down by materials (materials such as aluminum, argon, carbon, silicon, etc.) with high scatter cross sections in at least the high energy range.

The filter should reduce the gamma ray contamination, thermal neutron current and the fast neutron current contamination but fast neutrons should not be captured since the intensity of the irradiation beam will be very low; therefore fast neutrons must be moderated to remain a high intensity [4].

In general, the assembly filter design should thus provide sufficient shielding against fast neutrons and gamma rays while maintaining high epithermal neutron intensity [5].

Therefore the filter can be split into three parts contain materials to reduce the gamma radiation, materials to reduce the fast neutron current contamination and materials that reduce the thermal neutron current contamination. Every part has its own specific materials and should be placed in that typical order.

The material commonly used in BNCT filters in order to decrease the number of fast neutrons is Al or its compounds. According to recommendations of IAEA (International Atomic Energy Agency) for BNCT [6], the desirable minimum beam intensity would be 109 epithermal neutrons cm-2s-1.

Elements which are good thermal neutron absorbers are: He-3,

Citation: Salehi D, Sardari D, Salehi M (2012) Evaluation of Design Neutron Filters in BNCT. 1:537 doi:10.4172/scientificreports.537

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B-10, Li-6, N-14 and Cl. Since He-3 is very rare, with abundance of 0.000138% in the earth’s crust, and has a high potential as an energy source, it’s not very useful as a thermal neutron absorber [4].

According to the result good moderators, or scatterers, are materials with mass numbers compared to the mass number of a neutron.

Neutron filter/moderator design

The goal of reactor epithermal neutron filter conceptual design effort was to provide a low cost, simple filter design that is easy to fabricate and install while meeting or exceeding the nuclear performance requirements of the BNCT program. The epithermal neutron flux is defined as:

10

0.414( )

νϕ ϕ= ∫

keV

e neE dE

The neutron filter/moderator should have a high scattering cross section in the high energy range (fast neutrons) and a low cross section for the epithermal neutrons [7].

A design approach was selected that allows the filter to be pre-assembled to verify interfaces prior to installation at the reactor.

The filter will be assembled in layers to allow for easy handling while working in the empty thermal column. The filter can also be easily disassembled and reconfigured to allow for future adjustments in the nuclear performance. Most of the fast neutrons are moderated to epithermal and thermal energies as they scatter through the plates of aluminum and Teflon.

Since thermal neutrons are not desired at the surface of the patient, a material that absorbs thermal neutrons is also required in the filter.

Effective shaping materials should reduce fast neutron flux while enhancing epithermal flux. In other words, they allow for efficient slowing–down from the fast to the epithermal energies without removing these epithermal neutrons [8]. Good shaping materials are described as follows:

FluentalTM: FluentalTM is a patented material (i.e. metal and ceramic) for neutron moderation. It consists of a mixture of 30% aluminum (Al) (by weight), 69% aluminum fluoride (AlF3), and 1% lithium fluoride (LiF) [9] and is used as a moderator in some BNCT facilities (this material is an excellent moderator for accelerator-based neutron sources for BNCT) that was developed in Finland for reactor-based BNCT. Fluental was first suggested as a useful moderator material for an ABNS by Nigg et al. [10]. It was pioneered at LBL [11] as a moderator material for the 7Li(p,n)7Be reaction.

Since the main component of FluentalTM is aluminum and fluorine, energy deposition of the fast neutrons to the moderator in one collision becomes smaller rather than other materials examined. This result indicates that the quantity of FluentalTM is insufficient in such a limited moderator space to moderate fast neutrons [5].

Kim and Kim [8] on their work were reported that the ratio Σs, fast → epi /Σr, epi is the most important parameter identifying good moderators. In figure 1 the results has been collected. So Compounds containing fluorine are able to concentrate neutrons better in the epithermal energy (during slowing-down from higher energies).

Al and F and teflon compounds: Al and AlF3 are the most common moderator materials in beam-shaping assemblies for accelerator-based BNCT. However, their low density compared to FluentalTM is one of their main drawbacks-relatively thick Al and AlF3 are required for the

same neutron attenuation. On the other hand, these materials are often used because of their low cost compared to FluentalTM. Magnesium fluorine (MgF2) is also useful for producing neutrons in the epithermal energy range.

This is mainly due to strong, elastic-scattering cross section for fast neutrons and also these two elements have relatively small mass number. A recent publication highlights that moderators containing MgF2 produce neutron spectra appropriate for BNCT while at the same time minimizing fast neutrons. The result is relatively low radiation damage of surface tissue compared with others [8].

Combinations of Al followed by Al2O3 or AlF3 downstream, i.e. near the beam exit, are very efficient moderators, because the O and F cross-sections fill in the valleys between the energy resonance peaks of Al is studied by Jacob Jan de Boer [4].

His work shows that AlF3 is a better moderator for neutron energies less than 10 keV, where Al2O3 is a better moderator for fast neutron energies (higher than 10 keV).

The aluminum-teflon combination be employed because it produces good beam characteristics, and the materials are easily available and not too expensive [2].

The A1F3 material is somewhat more effective in reducing the fast neutron component of the spectrum for a given filter thickness.

Teflon (polytetrafluoroethylene; -[-CF2-CF2-]n-), which has a significantly greater resistance to radiation damage than one might first think, is an inexpensive, readily available, chemically stable, fluorine containing material, which in combination with aluminum, was postulated to be an effective neutron filtering material, provided that there is no unacceptable level of spectral degradation by elastic scattering from carbon.

Lithium and compounds: Natural lithium was selected as the best thermal neutron absorber because few gamma rays are released in the process (neutron capture in lithium-6 produces tritium).

The high thermal absorption cross-section of the 6Li isotope helps to remove thermal neutrons from clinical neutron beams. The result is lower thermal contamination and lower gamma contamination due to thermal neutron absorption.

The 6Li filter increases the average energy of the epithermal neutrons in the epithermal neutron beam. This filter allows the beam to be useful for effective BNCT treatment at greater depth in tissue [12].

Figure 1: Variation Macroscopic Cross Section of Some Candidate Moderator.

Citation: Salehi D, Sardari D, Salehi M (2012) Evaluation of Design Neutron Filters in BNCT. 1:537 doi:10.4172/scientificreports.537

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The first filter assembly tested consisted of bricks molded from a composite material developed by the Technical Research Centre of Finland fluentalTM (Al/AlF3/LiF) and the second assembly was constructed specifically for this experiment as a novel aluminum/Teflon (A1CF2) "sandwich" design. Combination of laminated 0.635 cm Teflon (CF2) and aluminum sheets was also tested as a fluorine-based filter for epithermal neutron production.

Jacob Jan de Boer [4] has estimated epithermal neutron flux and specific doses in air for fluental, Al/AlF3, Al/Al2O3, Al/graphite, Bi, Pb and AlF3Ti in different thickness filters.

Computation

An initial series of experiments was performed to measure the neutron production rate in a small volume of D2O driven by a bremsstrahlung source and to obtain information concerning the variation of the photoneutron source strength with incident electron energy within the energy range of interest.

This part, about foil activation calculation is studied by Mitchell [1].

6LiF with higher cross section for neutron caused reactions reduces the number of epithermal neutrons rather than the fast ones; therefore, it is not a suitable material for BNCT purposes [6]. 6LiF seems to be very promising, because it effectively captures neutrons of low energies, below 10 eV, and reduces the number of neutrons of energies above approx 10 keV.

Monte Carlo N-Particle (MCNP) calculations were also conducted to evaluate the performance of gadolinium as the thermal neutron absorber in the neutron filter. Based on these calculations, gadolinium is not a suitable thermal neutron absorber for this application [2].

Heavy water: The use of FluentalTM provides sufficient epithermal neutrons, but the fast neutron flux is approximately 2.5 times larger than that in the case of heavy water.

Tyminska [6] concludes that D2O is the best filter for BNCT purposes reducing effectively the number of fast neutrons and increasing the number of neutrons in epithermal range. When the dimensions of the moderator materials are restricted, the use of heavy water as the neutron moderator was found to be the most suitable among the materials they examined.

Although the beryllium moderator also has a lower value, it cannot be used because of the low epithermal neutron intensity.

Miyamaru and Murata [5] used heavy water, fluentalTM, carbon, and beryllium for investigation as moderators. The results indicated that the use of heavy water as a moderator was suitable for epithermal neutron production with higher flux and decreasing number of fast neutrons.

Figure 2 is comparison between epithermal neutron flux for a given neutron flux input per moderator thickness for some of up compounds [13].

Historical review of filter component selection7Li(p,n)7Be reaction with Fluental/PbF2 moderator/reflector

assembly, important early contributions to BNCT filter assembly design, were made at OSU [14,15] and MIT [16,17], where D2O, aluminum/D2O mixtures and BeO were investigated as moderator materials in combination with alumina, lithium carbonate, and lead reflectors.

In 2001, FluentalTM was considered for use at the University of Birmingham with a graphite reflector [18]. At OSU, the moderator/reflector assembly design has evolved (at least temporarily) with a Fluental moderator, which is surrounded by a PbF2 reflector [19].

ResultsResults of the calculations that are done by Blue and Yanch [20] are

presented below:

For this moderator/reflector material combination for a moderator, which is 30 cm thick axially and 31 cm in diameter and is surrounded by an annular reflector, which is radially 31 cm thick. The calculations assume that a 10 mA beam of 2.5 MeV protons is perpendicularly, and uniformly, incident on a 25 cm diameter lithium-7 target.

In other work that was done by Venhuizen [2] two different neutron filtering assemblies were used, both of which were designed to moderate and filter the photoneutron source emanating from the heavy water in a manner that produces a neutron spectrum suitable for epithermal neutron BNCT at the irradiation point.

Figure 2: Compared epithermal neutron flux per moderator thickness for AlF3, LiF, MgF2, BeO D2O, Fluental, CaF2.

Foil reaction 115In(n,n')115m In 115In(n,γ)116In 63Cu(n,γ)64Cu 197Au(n,γ)198Auupstream 1.0800×l0-19 1.0960×10-17 - -

1.8480×l0-17

downstream 7.4860×l0-23 4.6430×l0-19 1.5175×l0-20 5.7780×10-19

5.4320×l0-19

Table 1: Reaction rate data for 6 MeV A1/A1F3 filter.

Foil reaction 115In(n,n')115m In 115In(n,γ)116In 63Cu(n,γ)64Cuupstream 2.149×l0-19 2.000×10-17 -

downstream 1.0690×l0-21 5.572×l0-19 4.1870×l0-20

9.251×l0-19

Table 2: Reaction rate data for 6 MeV Al/Teflon filter.

Interaction measured Energy of maximum cross section

Gamma activation energy (keV) counted

115In(n,γ)116In 1 eV resonance, thermals 1293,1097,416115In(n,n')115m In 430 keV threshold (fast) 336197Au(n,γ)198Au 5 eV resonance 41163Cu(n,γ)64Cu 1 keV resonance 511 (positron)

Table 3: Activation foil interactions used from experimental data for presentation of final epithermal neutron beam spectrum at 6 MeV.

Citation: Salehi D, Sardari D, Salehi M (2012) Evaluation of Design Neutron Filters in BNCT. 1:537 doi:10.4172/scientificreports.537

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Activation foil assemblies were selected and both were irradiated on the downstream and upstream sides of the moderating material (Al/AlF3/LiF and Al/Teflon filters).

Those downstream foils were placed as close as possible in a circle along the beam centerline. Foils were not positioned radially in these experiments. The upstream foil package consisted solely of indium foils but allowed for a direct comparison of the neutron flux intensity on opposing sides of the filter along the beam axis.

A thermal neutron absorber is required in the filter sandwich region to reduce the number of thermal neutrons streaming toward the patient, and to eliminate many of the gamma rays produced by thermal neutron capture in aluminum. Approximately 2% by weight of natural lithium is required to provide the absorber. However, an issue remains how to get natural lithium in or on the aluminum plates. The fast neutron component was measured by recording the 336 keV lines of the indium foils which is produced by a threshold inelastic scattering interaction.

Each of the filtering materials was tested using a set of foils chosen

Figure 3: Drawing of A1/A1F3 filter experimental geometry set-up.

Figure 4: Inferred measured and calculated photoneutron spectra through 30cm (Al/AlF3/LiF) composite filtering assembly using multiple-material-foil method at an electron current of 5.99 µamps.

to characterize a specific section of the energy spectra through the filtering materials and the 6 MeV data presented for the Aluminum/Teflon material was analyzed using the stacked-foil method, while similar results for the A1F3 material were completed using the multiple-foil-material method.

Calculated and measured reaction rates for foil reactions of interest for non-stacked foil packets irradiated with a neutron source produced by bremsstrahlung from a 6 MeV electron beam with both filter materials are listed in tables 1 and 2 [1].

In order to remove the thermal neutrons from the beam, a cadmium layer can be placed downstream of the neutron filter/moderator and in order to increase the neutron flux, the filter/moderator can be enclosed by a lead shield thickness [4].

These results shown for the Aluminum/Teflon material include indium foil as well as copper foil data in the stack. The actual foil

Figure 5: Inferred measured and calculated photoneutron spectra through 30 cm Al/Teflon filtering assembly using stacked-foil method at an electron Current of 6.19 µamps.

Figure 6: Upstream measured and calculated broad-group neutron fluxes through Al/AlF3/LiF composite filter at an electron current of 5.99 μamps.

Citation: Salehi D, Sardari D, Salehi M (2012) Evaluation of Design Neutron Filters in BNCT. 1:537 doi:10.4172/scientificreports.537

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Figure 7: Downstream measured and calculated broad-group neutron fluxes Through AI/AIF3/LiF composite filter at an electron current of 5.99 μamps.

Figure 8: Upstream measured and calculated broad-group neutron fluxes through AI/Teflon filter at an electron current of 6.19 μamps.

Figure 9: Downstream measured and calculated broad-group neutron fluxes Through Al/Teflon filter at an electron current of 6.19 μamps.

Figure 10: Shift in neutron flux spectrum through an ALF3 filter.

interactions of interest used in the final presentation of proof principle data from experiments are shown in table 3 [1].

Data selected from these experiments is presented for both filter materials at the 6 MeV electron energy [1].

By looking at the responses within specific energy regions from various foils, the total neutron spectrum can be reconstructed (limited by the foil activation analysis errors and by the number of foil energies selected).

The work shows that if it is of interest to determine the total neutron flux associated with a specific energy region, it is possible to employ suitable foils, subtracting the thermal contribution by means of a cadmium cover. The experimental geometry was similar to that shown in figure 3 except with Al/Tefion in place of AI/AIF3 as the filter material.

The results of the filtered photoneutron experiments are shown

in the spectrum plots of figures 4 and 5. The calculated spectral data in these graphs are plotted at the logarithmic midpoint in each broad energy group. Calculated neutron spectra for both upstream and downstream locations are indicated along with the inferred neutron flux data, unfolded from the foil activation rates using a direct matrix unfolding method.

The histogram plots of figures 6-9 show the integrated broad-group neutron. Fluxes are function of energy upstream and downstream for both filter materials [1].

Optimum AlF3 filter length

As the length of the AlF3 filter was increased the neutron flux spectra (figure 10) showed that the fast neutron flux decreased relative to the epithermal flux. For an insight into the reason for this decrease, the attenuation of the various components of dose to tissue was

Citation: Salehi D, Sardari D, Salehi M (2012) Evaluation of Design Neutron Filters in BNCT. 1:537 doi:10.4172/scientificreports.537

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Figure 11: Components of dose a function of ALF3 Filter length at a depth of 3cm in phantom.

investigated as a function of AlF3 filter length; the neutron, gamma and 10B dose rates at a depth of 3 cm in tissue was investigated as a function of AlF3 filter length are shown in figure 11 [3].

ConclusionThe design of the moderator and filter assembly is driven by clinical

requirements and geared towards high quality epithermal neutron beams while maintaining a short treatment time. The epithermal neutron flux level at the patient position should be greater than 109 ncm-2s-1, while the fast neutron dose per unit incident flux at this position should be less than 5×10-11 cGy cm2, and the Gamma dose should be less than 10-11 cGy cm2. Neutrons with energies above 10 keV (fast neutrons) are very damaging to healthy tissue due to the recoil protons produced when neutrons undergo elastic scatter reactions with hydrogen. Therefore, it is important to design the neutron filter such that the relative intensity of these fast neutrons is low.

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