A shielding design for an accelerator-based neutron source for boron neutron capture therapy.

Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a (7)Li target, a target heat removal system (HRS), a moderator assembly, and a treatment room. The intent of this paper is to demonstrate the advantages of a shielded moderator assembly design, in terms of material requirements necessary to adequately protect radiation personnel located outside a treatment room for BNCT, over an unshielded moderator assembly design.

Common challenges and problems in clinical trials of boron neutron capture therapy of brain tumors.

Clinical trials for binary therapies, like boron neutron capture therapy (BNCT), pose a number of unique problems and challenges in design, performance, and interpretation of results. In neutron beam development, different groups use different optimization parameters, resulting in beams being considerably different from each other. The design, development, testing, execution of patient pharmacokinetics and the evaluation of results from these studies differ widely. Finally, the clinical trials involving patient treatments vary in many aspects such as their dose escalation strategies, treatment planning methodologies, and the reporting of data. The implications of these differences in the data accrued from these trials are discussed. The BNCT community needs to standardize each aspect of the design, implementation, and reporting of clinical trials so that the data can be used meaningfully.

An investigation on the use of removal-diffusion theory for BNCT treatment planning: a method for determining proper removal-diffusion parameters.

This paper outlines a method for determining proper removal-diffusion parameters to be used in removal-diffusion theory calculations for the purpose of BNCT treatment planning. Additionally, this paper demonstrates that, given the proper choice of removal-diffusion parameters, removal-diffusion theory may provide an accurate calculation technique for determining absorbed dose distributions for the purpose of BNCT treatment planning. For a four-group, one-dimensional calculation in water, this method was used to determine values for the neutron scattering cross sections, neutron removal cross sections, neutron diffusion coefficients, and extrapolation distances. These values were then used in a one-dimensional DIF3D calculation. The results of the DIF3D calculation showed a maximum deviation of 2.5% from a MCNP calculation performed for the same geometry.

Development and calculation of an energy dependent normal brain tissue neutron RBE for evaluating neutron fields for BNCT.

In Boron Neutron Capture Therapy (BNCT) of malignant brain tumors, the energy dependence of a clinically relevant Relative Biological Effectiveness (RBE) for epithermal neutrons, RBE(En), is important in neutron field design. In the first half of this paper, we present the development of an expression for the energy dependent normal-tissue RBE, RBE(En). We then calculate a reasonable estimate for RBE(En) for adult brain tissue. In the second half of the paper, two separate RBE expressions are developed, one for the RBE of the neutrons that interact in tissue via the 14N(n,p)14C reaction, denoted RBE(N), and one for the RBE of the neutrons which interact in tissue via the 1H(n,n')1H reaction, denoted RBE(H). The absorbed-dose-averaged values of these expressions are calculated for the neutron flux spectrum in phantom for the Brookhaven Medical Research Reactor (BMRR) epithermal neutron beam. The calculated values, [RBE(norm)N] = 3.4 and [RBE(norm)H] = 3.2, are within 6% of being equal, and support the use of equal values for RBEN and RBE(H) by researchers at Brookhaven National Laboratory (BNL). Finally, values of [RBE(norm)N] and [RBE(norm)H], along with the absorbed-dose-averaged RBE for brain, [RBE(norm)b], were calculated as a function of depth along the centerline of an ellipsoidal head phantom using flux spectra calculated for our Accelerator-Based Neutron Source (ABNS). These values remained essentially constant with depth, supporting the use of constant values for RBE, as is done at BNL.

Absorbed dose estimates to structures of the brain and head using a high-resolution voxel-based head phantom.

The purpose of this article is to demonstrate the viability of using a high-resolution 3-D head phantom in Monte Carlo N-Particle (MCNP) for boron neutron capture therapy (BNCT) structure dosimetry. This work describes a high-resolution voxel-based model of a human head and its use for calculating absorbed doses to the structures of the brain. The Zubal head phantom is a 3-D model of a human head that can be displayed and manipulated on a computer. Several changes were made to the original head phantom which now contains over 29 critical structures of the brain and head. The modified phantom is a 85 x 109 x 120 lattice of voxels, where each voxel is 2.2 x 2.2 x 1.4 mm3. This model was translated into MCNP lattice format. As a proof of principle study, two MCNP absorbed dose calculations were made (left and right lateral irradiations) using a uniformly distributed neutron disk source with an 1/E energy spectrum. Additionally, the results of these two calculations were combined to estimate the absorbed doses from a bilateral irradiation. Radiobiologically equivalent (RBE) doses were calculated for all structures and were normalized to 12.8 Gy-Eq. For a left lateral irradiation, the left motor cortex receives the limiting RBE dose. For a bilateral irradiation, the insula cortices receive the limiting dose. Among the nonencephalic structures, the parotid glands receive RBE doses that were within 15% of the limiting dose.

A comparison of neutron beams for BNCT based on in-phantom neutron field assessment parameters.

In this paper our in-phantom neutron field assessment parameters, T and DTumor, were used to evaluate several neutron sources for use in BNCT. Specifically, neutron fields from The Ohio State University (OSU) Accelerator-Based Neutron Source (ABNS) design, two alternative ABNS designs from the literature (the Al/AIF3-Al2O3 ABNS and the 7LiF-AI2O3 ABNS), a fission-convertor plate concept based on the 500-kW OSU Research Reactor (OSURR), and the Brookhaven Medical Research Reactor (BMRR) facility were evaluated. In order to facilitate a comparison of the various neutron fields, values of T and DTumor were calculated in a 14 cm x 14 cm x 14 cm lucite cube phantom located in the treatment port of each neutron source. All of the other relevant factors, such as phantom materials, kerma factors, and treatment parameters, were kept the same. The treatment times for the OSURR, the 7LiF-Al2O3 ABNS operating at a beam current of 10 mA, and the BMRR were calculated to be comparable and acceptable, with a treatment time per fraction of approximately 25 min for a four fraction treatment scheme. The treatment time per fraction for the OSU ABNS and the Al/AlF3-Al2O3 ABNS can be reduced to below 30 min per fraction for four fractions, if the proton beam current is made greater than approximately 20 mA. DTumor was calculated along the bean centerline for tumor depths in the phantom ranging from 0 to 14 cm. For tumor depths ranging from 0 to approximately 1.5 cm, the value of DTumor for the OSURR is largest, while for tumor depths ranging from 1.5 to approximately 14 cm, the value of DTumor for the OSU-ABNS is the largest.

Evaluation of scatter contribution from shielding materials used in scatter measurements for calibration range characterization.

The contribution of scatter from the shielding material used in scatter measurements at the Ohio Emergency Management Agency was evaluated. This was accomplished by comparing physical measurements with results generated by detectors modeled in the Monte Carlo shielding code, MCNP4B. The high purity germanium (HPGe) detector model was improved by using radiographs to more accurately represent the detector geometry, including a model of the lithium-contact transition region, and considering the effects of the non-uniform electric field in the germanium crystal. Photons scattered from the shielding material were found to contribute 4.25% of the total scatter recorded by the HPGe detector (thus contributing only 0.26% of the total dose) at a point 5 m from a 1.35 GBq (36.4 mCi) 137Cs source.

Predictions of a stochastic model of bone marrow cell survival in high dose rate radiation fields with arbitrary neutron to gamma-ray absorbed dose rate ratios.

In this paper, a stochastic model of cell survival, which was developed by Cotlet and Blue, based on the work of Jones, is extended to describe bone marrow cell survival in high dose rate radiation fields with arbitrary neutron to gamma-ray absorbed dose rate ratios. Mathematical formulas are obtained that describe the interaction of the neutron and gamma-ray components of the absorbed dose, for radiation fields with arbitrary neutron to gamma-ray dose rate ratios, for exposures of cells to various absorbed doses, at various high dose rates.

A stochastic model of radiation-induced bone marrow damage.

A stochastic model, based on consensus principles from radiation biology, is used to estimate bone-marrow stem cell pool survival (CFU-S and stroma cells) after irradiation. The dose response model consists of three coupled first order linear differential equations which quantitatively describe time dependent cellular damage, repair, and killing of red bone marrow cells. This system of differential equations is solved analytically through the use of a matrix approach for continuous and fractionated irradiations. The analytic solutions are confirmed through the dynamical solution of the model equations using SIMULINK. Rate coefficients describing the cellular processes of radiation damage and repair, extrapolated to humans from animal data sets and adjusted for neutron-gamma mixed fields, are employed in a SIMULINK analysis of criticality accidents. The results show that, for the time structures which may occur in criticality accidents, cell survival is established mainly by the average dose and dose rate.

Synthesis of 5-(carboranylalkylmercapto)-2'-deoxyuridines and 3-(carboranylalkyl)thymidines and their evaluation as substrates for human thymidine kinases 1 and 2.

Derivatives of thymidine containing o-carboranylalkyl groups at the N-3 position and derivatives of 2'-deoxyuridine containing o-carboranylalkylmercapto groups at the C-5 position were synthesized. The alkyl spacers consist of 4-8 methylene units. The synthesis of the former compounds required 3-4 reaction steps in up to 75% overall yield and that of the latter 9-10 reaction steps with significantly lower overall yield. Derivatives of thymidine substituted with carboranylalkyl substituents at the N-3 position and short spacers were phosphorylated by both recombinant and purified cytosolic thymidine kinase (TK1) to a relatively high degree. None of the tested 2'-deoxyuridine derivatives possessing carboranyl substituents at the C-5 position were phosphorylated by either recombinant or purified TK1. The amounts of phosphorylation product detected for some of the C-5-substituted nucleosides with recombinant mitochondrial thymidine kinase (TK2) were low but significant and decreased with increasing lengths of the alkyl spacer. The data obtained in this study do not seem to support the tether concept applied in the synthesis of the new C-5- and N-3-substituted carboranyl nucleosides intended to reduce possible steric interference in the binding of carboranyl nucleosides with deoxynucleoside kinases. Instead, it appeared that a closer proximity of the bulky carborane moiety to the nucleoside scaffold resulted in better substrate characteristics.

Boron neutron capture therapy of brain tumors: an emerging therapeutic modality.

Boron neutron capture therapy (BNCT) is based on the nuclear reaction that occurs when boron-10, a stable isotope, is irradiated with low-energy thermal neutrons to yield alpha particles and recoiling lithium-7 nuclei. For BNCT to be successful, a large number of 10B atoms must be localized on or preferably within neoplastic cells, and a sufficient number of thermal neutrons must be absorbed by the 10B atoms to sustain a lethal 10B (n, alpha) lithium-7 reaction. There is a growing interest in using BNCT in combination with surgery to treat patients with high-grade gliomas and possibly metastatic brain tumors. The present review covers the biological and radiobiological considerations on which BNCT is based, boron-containing low- and high-molecular weight delivery agents, neutron sources, clinical studies, and future areas of research. Two boron compounds currently are being used clinically, sodium borocaptate and boronophenylalanine, and a number of new delivery agents are under investigation, including boronated porphyrins, nucleosides, amino acids, polyamines, monoclonal and bispecific antibodies, liposomes, and epidermal growth factor. These are discussed, as is optimization of their delivery. Nuclear reactors currently are the only source of neutrons for BNCT, and the fission reaction within the core produces a mixture of lower energy thermal and epithermal neutrons, fast or high-energy neutrons, and gamma-rays. Although thermal neutron beams have been used clinically in Japan to treat patients with brain tumors and cutaneous melanomas, epithermal neutron beams now are being used in the United States and Europe because of their superior tissue-penetrating properties. Currently, there are clinical trials in progress in the United States, Europe, and Japan using a combination of debulking surgery and then BNCT to treat patients with glioblastomas. The American and European studies are Phase I trials using boronophenylalanine and sodium borocaptate, respectively, as capture agents, and the Japanese trial is a Phase II study. Boron compound and neutron dose escalation studies are planned, and these could lead to Phase II and possibly to randomized Phase III clinical trials that should provide data regarding therapeutic efficacy.

The rationale and requirements for the development of boron neutron capture therapy of brain tumors.

The dismal clinical results in the treatment of glioblastoma multiforme despite aggressive surgery, conventional radiotherapy, and chemotherapy, either alone or in combination has led to the development of alternative therapeutic modalities. Among these is boron neutron capture therapy (BNCT). This binary system is based upon two key requirements: (1) the development and use of neutron beams from nuclear reactors or other sources with the capability for delivering high fluxes of thermal neutrons at depths sufficient to reach all tumor foci, and (2) the development and synthesis of boron compounds that can penetrate the normal bloodbrain barrier, selectively target neoplastic cells, and persist therein for suitable periods of time prior to irradiation. The earlier clinical failures with BNCT related directly to the lack of tissue penetration by neutron beams and to boron compounds that showed little specificity for and low retention by tumor cells, while attaining high concentrations in blood. Progress has been made both in neutron beam and compound development, but it remains to be determined whether these are sufficient to improve therapeutic outcomes by BNCT in comparison with current therapeutic regimens for the treatment of malignant gliomas.

Shielding design of a treatment room for an accelerator-based epithermal neutron irradiation facility for BNCT.

Protecting the facility personnel and the general public from radiation exposure is a primary safety concern of an accelerator-based epithermal neutron irradiation facility. This work makes an attempt at answering the questions "How much?" and "What kind?" of shielding will meet the occupational limits of such a facility. Shielding effectiveness is compared for ordinary and barytes concretes in combination with and without borated polyethylene. A calculational model was developed of a treatment room , patient "scatterer," and the epithermal neutron beam. The Monte Carlo code, MCNP, was used to compute the total effective dose equivalent rates at specific points of interest outside of the treatment room. A conservative occupational effective dose rate limit of 0.01 mSv h-1 was the guideline for this study. Conservative Monte Carlo calculations show that constructing the treatment room walls with 1.5 m of ordinary concrete, 1.2 m of barytes concrete, 1.0 m of ordinary concrete preceded by 10 cm of 5% boron-polyethylene, or 0.8 m of barytes concrete preceded by 10 cm of 5% boron-polyethylene will adequately protect facility personnel.

An in-situ method to measure a soil's undisturbed pore gas radon concentration, diffusion length for radon and air filled porosity.

Previous work has shown that work has shown that for soils of insignificant permeability (soils with permeability less than about 10(-12) m2) the important soil parameters for characterizing radon mobility in the soil are the soil's steady-state pore gas random concentration at depth (Cs), the soil's bulk diffusion length for radon (L) and the soil's air filled porosity (pa). Existing methods to measure these parameters have been based wholly or in part on measurements of soil samples taken to a laboratory for analysis. The drawbacks of this approach are twofold: (1) since soil structure can be quite heterogeneous, the sample may not have characteristics indicative of the site as a whole, and (2) since the parameters are dependent on soil structure and the soil structure of the sample may be changed in th e process of acquiring the sample, one may be changing the parameters that one is trying to measure. These problems can be avoided by using a totally in-situ method to measure Cs, L, and pa. This paper describes a totally in-situ method for simultaneously measuring the important soil parameters, based on measurements of the radon concentration as a function of time for the gas in a cavity in the soil.

Steady-state response of a charcoal bed to radon in flowing air with water vapor.

Previously we have developed a mathematical model of radon adsorption in active air with water vapor on small U.S. Environmental Protection Agency charcoal canisters that are used for environmental measurements of radon. The purpose of this paper is to extend this mathematical model to describe the adsorption of radon by large charcoal beds with radon-laden air flowing through them. The resulting model equations are solved analytically to predict the steady-state adsorption of radon by such beds.

An expression for the RBE of neutrons as a function of neutron energy.

The goal of this paper is to develop a relationship between a neutron RBE and neutron energy, En, which can be used to design neutron sources for BNCT. In an earlier calculation of a neutron RBE as a function of En, we approximated the contribution to a total neutron RBE, RBEt (En), arising from 14N(n,p)14C reactions. In this paper, we recalculate RBEt (En), accounting more exactly for the contribution to RBEt (En) from 14N(n,p)14C reactions.

Dose prescription in boron neutron capture therapy.

PURPOSE: The purpose of this paper is to address some aspects of the many considerations that need to go into a dose prescription in boron neutron capture therapy (BNCT) for brain tumors; and to describe some methods to incorporate knowledge from animal studies and other experiments into the process of dose prescription. MATERIALS AND METHODS: Previously, an algorithm to estimate the normal tissue tolerance to mixed high and low linear energy transfer (LET) radiations in BNCT was proposed. We have developed mathematical formulations and computational methods to represent this algorithm. Generalized models to fit the central axis dose rate components for an epithermal neutron field were also developed. These formulations and beam fitting models were programmed into spreadsheets to simulate two treatment techniques which are expected to be used in BCNT: a two-field bilateral scheme and a single-field treatment scheme. Parameters in these spreadsheets can be varied to represent the fractionation scheme used, the 10B microdistribution in normal tissue, and the ratio of 10B in tumor to normal tissue. Most of these factors have to be determined for a given neutron field and 10B compound combination from large animal studies. The spreadsheets have been programmed to integrate all of the treatment-related information and calculate the location along the central axis where the normal tissue tolerance is exceeded first. This information is then used to compute the maximum treatment time allowable and the maximum tumor dose that may be delivered for a given BNCT treatment. RESULTS AND CONCLUSION: The effect of different treatment variables on the treatment time and tumor dose has been shown to be very significant. It has also been shown that the location of Dmax shifts significantly, depending on some of the treatment variables--mainly the fractionation scheme used. These results further emphasize the fact that dose prescription in BNCT is very complicated and nonintuitive. The physician prescribing the dose would need to rely on some method, like the one developed here, to come up with an appropriate dose prescription.

Effect of head phantom size on 10B and 1H[n,gamma]2H dose distributions for a broad field accelerator epithermal neutron source for BNCT.

The effect of head phantom size on the 10B and 1H[n,gamma]2H dose distributions for a broad epithermal neutron radiation field generated by an accelerator-based epithermal neutron source for boron neutron capture therapy (BNCT) have been studied. Also two techniques for calculating the absorbed gamma dose from a measured gamma-ray source distribution are compared: a Monte Carlo technique, which is well accepted in the BNCT community, and a Point Kernel technique. The count-rate distribution in the central plane of three rectangular parallelopiped head water phantoms irradiated with an epithermal neutron field was measured with a boron trifluoride (BF3) detector. This epithermal neutron field was produced at the Ohio State University Van de Graaff Accelerator Facility. The 10B absorbed dose and the gamma-ray source have the same distribution in the head phantom as the BF3 count-rate distribution. The absorbed gamma dose from the measured source distribution was calculated using MCNP, a Monte Carlo code, and QAD-CGGP, a Point Kernel code. The most pronounced effect of phantom size on 10B absorbed dose was on the dose rate at the depth of maximum dose, dmax. An increase in dose rate at dmax was observed with a decrease in phantom size, the dose rate in the smallest phantom being larger by a factor of 1.4 than the dose rate in the largest phantom. Also, dmax for the phantoms shifted deeper with a decrease in phantom dimensions. The shift between the largest and the smallest phantoms was 6 mm. Finally, the smaller phantoms had lower entrance 10B dose as a percent of the dose at dmax, or better skin sparing. Our calculations for the gamma dose show that a Point Kernel technique can be used to calculate the dose distribution as accurately as a Monte Carlo technique, in much shorter computation times.

A model for radon gas adsorption on charcoal for open-faced canisters in an active environment.

It was previously suggested that the calibration factor (CF) relating the detected activity (A) of the radon decay products 214Bi and 214Pb within a charcoal canister to the radon concentration (CR) in air be defined as CF = A/(E x DF x CR), where E is the counting efficiency of the detector for 214Bi and 214Pb gamma rays, and DF is a factor accounting for decay during the time (tD) from the end of the exposure until the canister is counted; i.e., DF = exp(-lambda RPtD), where lambda RP is the physical decay constant for radon. With CF defined as above, a kinetic model for the adsorption of radon gas on charcoal in humid air predicts that CF (in m3) can be written for CR in Bq m-3, A in cpm, and E in cpm Bq-1 as CF = alpha[1-exp(-delta t)] + beta[1 - exp(-delta t)]M(t), where alpha, beta, and delta are free parameters, t is the canister exposure time in hours, and M(t) is the canister water mass gain in grams in time t. For CF data for the U.S. Environmental Protection Agency's open-faced canisters in an active (with respect to air flow) environmental chamber, delta = (0.046 +/- 0.002)h-1, and exp(-delta t) much less than 1 for t greater than or equal to 3 d. Also, alpha = (0.342 +/- 0.002)m3, and beta = -(0.0123 +/- 0.0001)m3g-1.

An experimental study of the moderator assembly for a low-energy proton accelerator neutron irradiation facility for BNCT.

An accelerator-based neutron irradiation facility (ANIF), which has been proposed for BNCT, is based on a 2.5-MeV proton beam bombarding a thick lithium target. Neutrons which are emitted from the lithium target are too energetic for BNCT and must be moderated. A calculational study, which was done previously on the moderator assembly for an ANIF, shows that, with an optimized moderator assembly, an ANIF can produce a neutron flux which has quality and intensity sufficient for BNCT. In order to verify our previous calculational study, a lithium target and a non-optimized moderator assembly (a cylindrical tank of D2O) have been constructed and tested at the Ohio State University Van de Graaff proton accelerator. The neutron spectrum was measured for neutrons emerging from the moderator assembly. The measured neutron spectrum agrees reasonably well with that obtained from Monte Carlo calculations, except for neutrons with energies above 100 keV. For those neutrons, the measured spectrum is lower by a factor of two than the calculated one. In addition to the neutron spectrum measurement, the boron-10 absorbed dose was measured on the axis of the neutron field in a 20 cm x 20 cm x 20 cm water phantom, and the result agrees quite well with that obtained from calculation. This experiment confirms that the calculated optimized moderator assembly, consisting of a 22.5-cm thick, 25-cm diameter cylinder of beryllia (BeO) surrounded by a 30-cm thick jacket of alumina (Al2O3), produces an epithermal neutron flux of 3.12 x 10(7) n/cm2-s per mA of protons. For an accelerator delivering 30 mA of 2.5-MeV protons, the irradiation time for a single-session treatment can be as short as 50 minutes. The calculated ratio of absorbed neutron dose to fluence for the optimized moderator assembly is 4.9 x 10(-11) cGy-cm2/n, which is equal to that of a 5-keV neutron beam. Our experimental measurements indicate that the ratio of absorbed neutron dose to fluence may in fact be lower (better) than calculated.