Application of the new MultiTrans radiation transport code in BNCT dose planning

Abstract

Dose planning in boron neutron capture therapy (BNCT) is a complex problem and requires sophisticated numerical methods. In the framework of the Finnish BNCT project, new deterministic three-dimensional radiation transport code MultiTrans SP3 has been developed at VTT Chemical Technology, based on a novel application of the tree multigrid technique. To test the applicability of this new code in a realistic BNCT dose planning problem, cylindrical PMMA (polymethyl-methacrylate) phantom was chosen as a benchmark case. It is a convenient benchmark, as it has been modeled by several different codes, including well-known DORT and MCNP. Extensive measured data also exist. In this paper, a comparison of the new MultiTrans SP3 code with other methods is presented for the PMMA phantom case. Results show that the total neutron dose rate to ICRU adult brain calculated by the MultiTrans SP3 code differs less than 4% in 2 cm depth in phantom (in thermal maximum) from the DORT calculation. Results also show that the calculated 197Au(n,gamma) and 55Mn(n,gamma) reaction rates in 2 cm depth in phantom differ less than 4% and 1% from the measured values, respectively. However, the photon dose calculated by the MultiTrans SP3 code seems to be incorrect in this PMMA phantom case, which requires further studying. As expected, the deterministic MultiTrans SP3 code is over an order of magnitude faster than stochastic Monte Carlo codes (with similar resolution), thus providing a very efficient tool for BNCT dose planning.

Light transport in biological tissue based on the simplified spherical harmonics equations

“…The SP N method was originally proposed by Gelbard [20,21] for neutron transport, but a thorough theoretical foundation was laid out by Larsen and co-workers only during the 1990's [22][23][24][25][26]. The SP N equations have also been applied to other problems in particle transport and heat radiation transfer [27][28][29][30][31][32][33]. However, there are distinct differences between the formulations of the SP N equations in nuclear sciences and tissues optics.…”

“…The SP N methods provide a convenient framework for modeling collimated boundary sources by using a direction-dependent source function S(X) in Eqs. (28). Including this physics will significantly increase the effect of boundary layers on the problem and should measurably improve the SP N solutions in comparison to the diffusion solutions.…”

Boron neutron capture therapy (BNCT) in Finland: Technological and physical prospects after 20 years of experiences

“…[23,102], the THORplan developed at Tsing Hua University in Taiwan [103], and the JAEA computational dosimetry system (JCDS) developed at Japan Atomic Energy Agency (JAEA). The MC methods are often very time consuming, which has led to interest in the application of faster deterministic methods [23], such as the simplified P n approach [104].…”

On the development of computational tools for the design of beam assemblies for boron neutron capture therapy

“…where ψ + R−K and ψ − R−K are (N/2)th-dimensional column matrices whose entries are the incident and backscattered angular fluxes at x = x R−K , respectively, and ψ + R is column matrix (14). Then, there exist corresponding matrix operators B R−K,R and T R−K,R for the diffuse reflection by and transmission through the (K + 1) layers (R − K), .…”

“…To cope with these tasks, a source-dependent beam shaping/filtering assembly (a block-type sandwich structure surrounded by a neutron reflector) is designed, optimized, fabricated and carefully tested before proper utilisation [10,11]. Particularly at the early design stage, when a variety of potential configurations for the assembly are investigated, simplified models for the trial assembly configurations are defined and, for each configuration, a beam transport problem is solved with a deterministic transport code (often an adaptation of a general-purpose nuclear reactor/radiation shielding code) in order to get estimates for the energy-angle distribution of neutrons leaving the assembly through the opposite (patient) side [9,[12][13][14][15][16][17] (it is important to note though that Monte Carlo [18,19] is the method of choice in BNCT optimization, dosimetry and treatment planning studies). As the number of configurations is high at the early design stage, the task of performing transport computations at this stage is burdensome and costly, even using efficient deterministic transport codes based on the well-known discrete ordinates formulation of neutron transport theory [9,12,13,[15][16][17].…”