Performance Characteristics of the MIT Epithermal Neutron Irradiation Facility

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Performance Characteristics of the MIT Epithermal Neutron Irradiation Facility

  1. 1. INSTITUTE OF PHYSICS PUBLISHING PHYSICS IN MEDICINE AND BIOLOGY Phys. Med. Biol. 48 (2003) 943–958 PII: S0031-9155(03)56482-X Performance characteristics of the MIT fission converter based epithermal neutron beam K J Riley, P J Binns and O K Harling Nuclear Reactor Laboratory and Department of Nuclear Engineering, Massachusetts Institute of Technology, 138 Albany Street, Cambridge, MA 02139, USA E-mail: flavor@mit.edu Received 19 November 2002 Published 18 March 2003 Online at stacks.iop.org/PMB/48/943 Abstract A pre-clinical characterization of the first fission converter based epithermal neutron beam (FCB) designed for boron neutron capture therapy (BNCT) has been performed. Calculated design parameters describing the physical performance of the aluminium and Teflon R filtered beam were confirmed from neutron fluence and absorbed dose rate measurements performed with activation foils and paired ionization chambers. The facility currently provides an epithermal neutron flux of 4.6 × 109 n cm−2 s−1 in-air at the patient position that makes it the most intense BNCT source in the world. This epithermal neutron flux is accompanied by very low specific photon and fast neutron absorbed doses of 3.5 ± 0.5 and 1.4 ± 0.2 × 10−13 Gy cm2, respectively. A therapeutic dose rate of 1.7 RBE Gy min−1 is achievable at the advantage depth of 97 mm when boronated phenylalanine (BPA) is used as the delivery agent, giving an average therapeutic ratio of 5.7. In clinical trials of normal tissue tolerance when using the FCB, the effective prescribed dose is due principally to neutron interactions with the nonselectively absorbed BPA present in brain. If an advanced compound is considered, the dose to brain would instead be predominately from the photon kerma induced by thermal neutron capture in hydrogen and advantage parameters of 0.88 Gy min−1, 121 mm and 10.8 would be realized for the therapeutic dose rate, advantage depth and therapeutic ratio, respectively. This study confirms the success of a new approach to producing a high intensity, high purity epithermal neutron source that attains near optimal physical performance and which is well suited to exploit the next generation of boron delivery agents. 1. Introduction Research to try and establish boron neutron capture therapy (BNCT) as a viable treatment modality for cancer has to date emphasized using reactor produced beams to irradiate 0031-9155/03/070943+16$30.00 © 2003 IOP Publishing Ltd Printed in the UK 943
  2. 2. 944 K J Riley et al patients suffering from cerebral glioma, a particularly refractory type of malignancy. Early investigations were performed with beams of thermal neutrons that, because they lacked penetration, could only be used with therapeutic intent in conjunction with intra-operative procedures. Such surgical interventions can be avoided if better penetrating beams of epithermal neutrons are utilized instead to irradiate deep-seated lesions. Developments in beam technology during the last decade enabled epithermal neutrons to be extracted from reactor cores with sufficient intensity and quality suitable for BNCT. The first such clinical irradiations with epithermal neutrons were conducted in the USA at the research reactor of the Massachusetts Institute of Technology (MITR) and the Brookhaven Medical Research Reactor (BMRR) (Rogus et al 1994, Liu et al 1996, Busse et al 1997, 2002, Chanana et al 1999). This initiative prompted a number of other dedicated facilities to be commissioned around the world that are now actively pursuing clinical trials with epithermal neutrons. Dose escalation studies were initially performed with the M67 epithermal beam at MIT and as clinical experience was gained the limitations of this facility became more evident. In particular, the low flux of epithermal neutrons resulted in protracted irradiation times of several hours for a single field and positioning the patient against the ceiling in the vertical beam line was often difficult. Although the M67 beam line was useful for initiating brain tolerance studies, it soon became apparent that a more advanced facility would be needed for subsequent clinical trials and possible routine therapy applications. Further modifications to increase the intensity of the M67 beam were not technically feasible and as an alternative a completely new facility was considered that would satisfy clinical needs without significant compromise in the foreseeable future. The design specifications of the new facility included a copious source of epithermal neutrons to provide a high intensity beam with minimal fast neutron, slow neutron or photon contamination. The dose distributions produced by the beam in target volumes should be governed predominantly by the uptake characteristics of the administered boron delivery agent and the beam intensity should allow for short irradiation times comparable to those in conventional therapy. To help satisfy these requirements, an upper limit for the allowable beam contamination was considered at approximately an order of magnitude below the 2 × 10−12 Gy cm2 inherent background kerma that is inevitably produced in tissue by neutron capture reactions with nitrogen and hydrogen. The ability to readily increase flux production further was also desirable for possible future optimization of the beam by using additional filtration or collimation. A forward directed beam with variable collimation and flexible patient positioning to permit irradiation of disease sites anywhere on the body was also considered advantageous. Other features to be incorporated for operational convenience and safety were an automated dose monitoring and control system together with beam line shutters and a well-shielded medical room that would be easily accessible with the reactor at full power. The design of the new facility should also enable independent operation of the beam without affecting other user applications at the multipurpose MITR. Extensive feasibility studies were performed and a neutronic design based upon the use of a fission converter source at the MITR appeared to adequately fulfil all of the desired criteria (Sakamoto et al 1999, Kiger et al 1999). Design calculations identified aluminium and Teflon R as an efficient and inexpensive neutron filter-moderator and subsequent irradiation tests demonstrated that Teflon R would perform satisfactorily in the beam line over the expected lifetime of the FCB (Harling et al 2002a). Following successful funding applications, construction of the fission converter beam line began in the autumn of 1997 and the new facility was commissioned three years later (Harling et al 2002b). As part of the ensuing commissioning process an extensive dosimetric evaluation was needed to assess the actual performance of the fission converter produced beam and to provide the necessary data for
  3. 3. Performance characteristics of the MIT FCB 945 ◦ Figure 1. Plan view of the new fission converter based beam line at MIT. The beam line consists of an aluminium and Teflon R filter-moderator as well as cadmium and lead filters. Beam delivery is controlled with three in-line shutters that are interspersed along the length of the beam line. the clinical trials that have recently resumed at MIT. This included measuring the neutron flux together with the neutron and photon absorbed dose components in-air as well as in- phantom. Thermal flux in-phantom was determined from the activation of gold foils while dual ionization chambers with walls constructed of A-150 tissue equivalent plastic and graphite were used to measure the fast neutron and photon absorbed dose components, respectively. Further investigations were also performed using a fission chamber to verify symmetry of the epithermal neutron distribution in the collimated field. The results of these measurements and comparison with the specified design goals for the new facility are described. To illustrate the performance of the FCB under conditions pertinent to clinical irradiations, figures of merit were determined from the biologically weighted depth dose profiles for both a currently approved boron compound and an advanced delivery agent that offers greater selectivity. 2. Method 2.1. Description of the facility The new epithermal neutron irradiation facility is housed in the experimental hall of the MITR and can operate simultaneously with other user applications. A plan view of the beam line is shown in figure 1. Thermal neutrons from the reactor core impinge upon a sub-critical array of fuel elements housed in a separate, well-shielded vessel, located outside the reflector region. In its current configuration, the converter contains ten partially burned MITR fuel elements (average loading of 338 g 235U per element) that are cooled by forced convection of heavy water and produce a power of 83 kW with the reactor at 5 MW. The converter is designed and licensed to operate at a power of 250 kW that could be achieved by loading with a full complement of fresh MITR fuel and upgrading the reactor systems to operate at 10 MW. A re-licensing request for the MITR to operate at 6 MW is currently pending with the Nuclear Regulatory Commission. A shielded horizontal beam line 2.5 m long directs neutrons from the converter to the treatment room. The beam line itself consists of a series of aluminium
  4. 4. 946 K J Riley et al (0.81 m) and Teflon R (0.13 m) filter-moderators, a cadmium (0.5 mm) thermal neutron filter, a lead (60 mm) photon shield and a large conical collimator with lead walls 0.15 m thick. The aluminium and Teflon R filter moderators were chosen to tailor the fission spectrum to an epithermal energy distribution with an average neutron energy of approximately 2 keV. This energy distribution should result in a thermal neutron maximum at a depth in tissue of between 20 and 30 mm. Previous computational studies using ideal, mono-energetic neutron beams showed that neutrons with energies between approximately 100 eV and 20 keV possess the best depth dose characteristics for BNCT (Yanch and Harling 1993, Bisceglie et al 2000). Neutrons above or below this energy range provide more incidental dose without improving the build-up of thermal flux at depth. Beam delivery is controlled with three in-line shutters interspersed along the length of the beam line. The first of these is the converter control shutter (CCS) located on the reactor core side of the converter. The CCS regulates the fission rate produced in the converter by shielding the resident fuel from thermal neutrons that arrive from the reflector region. Lowering the CCS to the fully closed position reduces converter power by two orders of magnitude. Downstream from the fuel and closer to the patient position, is a 0.68 m long tank located in the lead collimator that when filled with light water attenuates neutrons. During an irradiation, the water is emptied and replaced by helium gas at ambient pressure. Nearest to the patient is the fast acting mechanical shutter that effectively turns the beam on and off for therapy. This shutter is a large sliding rectangular block that completely closes a section of the collimator in 8 s with a 0.20 m thickness of borated (100 mg cm−3 of 10B) high density concrete and 0.20 m of lead. Each shutter acts independently and all three completely close in under 3 min to allow medical staff unrestricted access to the new treatment room while the reactor is at full power. At the end of the beam line is the patient collimator that extends beyond the shielding wall and into the newly built medical room. The collimator allows easy patient positioning for irradiation of any envisaged disease site and is constructed of lead spheres (7.8 mm diameter) cast in epoxy resin loaded with either lithium fluoride (95% enriched 6Li) or boron carbide. Augmenting the collimator with additional end pieces provides variable circular apertures of 80, 100, 120 and 160 mm diameter that increase the distance from the shielding wall to the patient from 0.24 to 0.32 m. The neutron beam is continuously monitored with four pulse mode fission counters that are located at the periphery of the beam inside the patient collimator. The signal outputs from these counters are fed to NIM electronics that interface with the dose monitoring system consisting of two redundant industrial programmable logic controllers (PLCs) which operate in parallel. Beam delivery during therapy is controlled and monitored by the PLCs that signal the automatic opening and closing of all shutters to give the specified treatment prescription (Riley et al 2002). Data from the monitoring system are logged and instantly displayed on a separate computer equipped with Lookout R data acquisition software. 2.2. Computations The fission converter beam line was designed (Sakamoto et al 1999, Kiger et al 1999, Riley 2001) using the Monte Carlo code MCNP 4A (Briesmeister 1993). Beam characteristics were calculated at the patient position to give the thermal and epithermal neutron flux as well as the fast neutron and photon absorbed dose components. The computer model was driven by a surface source of neutrons that was written at the edge of the graphite reflector region, directly in front of the converter fuel elements (Redmond et al 1994). A new surface source was calculated at the end of the beam line closest to the patient (for each of the four available
  5. 5. Performance characteristics of the MIT FCB 947 collimator apertures) using energy-dependent weight windows as a variance reduction scheme to obtain the required statistical precision. A single group of photon weight windows was also used to provide an accurate representation of the prompt photon activity produced in the beam line. The absorbed dose rates were calculated in-air at the patient position by convolving surface flux tallies with appropriate photon and neutron kerma coefficients (ICRU 2000). A spatial profile of the neutron flux was calculated by tallying in concentric rings with increasing diameters of 20 mm. This width increment was chosen to match the active length of the detector used for measurements (Ali 2001). This profile, along with a similar calculation for photons, revealed that values in the centre of the beam (where measurements are performed) were no different than those averaged over the entire aperture, within the predicted uncertainties. The in-air neutron energy distribution was also calculated by segmenting the surface tally into the desired energy bins. The surface source for each aperture, containing both photons and neutrons, were in turn transported into an exact model of the water-filled ellipsoidal phantom in which measurements were performed. Neutron fluxes and dose rates were again tallied both at the incident surface of the phantom and in cylindrical cells of 12 mm diameter and 5 mm height, along the central axis of the phantom. The detector volumes for the corresponding measurements are smaller than those used in the calculations. The error induced by this discrepancy is, however, within the predicted uncertainties as the detectors were each placed in turn at the geometric centre of the respective tally volume, across which the dose gradient was linear. The neutron energy distribution was tallied as a function of depth in the water-filled phantom. All calculations are normalized to an MITR-II power of 5 MW that corresponds to the fission converter operating at 83 kW with the current fuel loading. 2.3. Measurements The fission rate in the converter is governed by the reactor operating power and can be independently controlled by the vertically opening CCS. The opening height of the CCS is continuously adjustable to provide the desired beam intensity. To check beam uniformity as a function of the CCS opening position, cross-plane profiles of the relative epithermal neutron intensity were obtained by scanning a fission chamber operated in pulse mode across the aperture at the exit of the collimator in the vertical direction. The small volume cylindrical chamber, with dimensions of 5 mm diameter and an active length of 19 mm, was irradiated side on. Count rates of pulses above a discriminator setting to reject alpha decay and photon events were registered using a single channel analyser and scaler. Although the fission chamber is sensitive to neutrons of all energies, approximately 85% of the fission events produced in the counter are attributed to neutrons in the epithermal (1 eV–10 keV) energy range (Wilson 2001). The epicadmium flux was measured in-air at the patient position using 0.05 mm thick gold foils and 0.5 mm thick cadmium covers for each of the four aperture sizes. Values of 1.56 × 10−21 cm2 for the (n, γ ) resonance integral in 197Au and 0.28 for the self-shielding factor of the foils were used (ASTM 1998). Calculations with the MCNP beam model showed that less than 1% of the measured activity induced in the Au was due to neutrons with energies greater than 10 keV and that the epicadmium and epithermal flux were essentially equivalent. Absorbed dose rates were measured on the central axis of the beam both free in-air, and at different depths in an ellipsoidal water phantom with major axes of 136, 196 and 166 mm (Harling et al 1995). A lateral irradiation of the brain was simulated by the Snyder-based phantom positioned with the shortest axis on the beam centre line. The water-filled phantom consists of a 2.5 mm shell constructed from quartz fibre and acrylic casting resin mounted to a
  6. 6. 948 K J Riley et al base of acrylic plates. Butyrate tubes 13 mm in outer diameter are inserted through a number of ports in the base to allow ionization chambers and gold foils to be positioned anywhere along its length. To avoid perturbing the thermal neutron flux in the vicinity of the gold foils during irradiations the tubes are filled with water. The absorbed dose rates arising from thermal neutrons captured in boron and nitrogen are calculated from the 2200 m s−1 neutron flux obtained from the activation of bare and cadmium covered gold foils (Rogus et al 1994, ASTM 1998). The measured 2200 m s−1 flux is multiplied by the appropriate weight fraction in the material of interest and kerma coefficients of 8.66 × 10−8 and 7.88 × 10−12 Gy cm2 for boron and nitrogen, respectively. Depth dose profiles were calculated assuming boron concentrations of 18 and 65 µg g−1 in normal brain and diffuse tumour tissue, respectively, that represent the characteristic uptakes observed using boronated phenylalanine (BPA) in earlier clinical trials at MIT (Kiger et al 2001). Additional profiles were determined for an advanced porphyrin compound, assuming 0.65 µg g−1 in brain and 65 µg g−1 in tumour (Miura et al 2001). The measurements and calculations reported assume no boron in the water of the phantom. The influence of boron has been quantified (Ye 1999) and is accurately accounted for in the treatment plans used during clinical irradiations (Zamenhof et al 1996). A nitrogen concentration of 3.5% by weight is assumed for both soft muscle tissue and A-150 plastic while 2.2% is assumed for brain (ICRU 1989). Biologically weighted depth dose profiles were obtained by applying RBE values of 1.0 for photons and 3.2 for thermal and fast neutrons. Differences in the effective micro- distribution of the boron delivered by BPA in both tissue and tumour were also accommodated using cRBE factors of 1.35 for normal brain and 3.8 for tumour (Coderre et al 1993). In the absence of information about the boron micro-distribution of porphyrin compounds, a cRBE of 3.8 was assumed for both normal brain and tumour. The neutron and photon absorbed dose rates were measured in the mixed radiation field using the dual ionization chamber technique (Attix 1986) with commercial counters (IC-18s Far West Technology) that have active volumes of 0.1 ml. The neutron sensitive chamber is constructed with A-150 plastic walls and was flushed with methane-based tissue equivalent gas (64.4% CH4, 32.4% CO2 and 3.2% N2) at a flow rate of 20 ml min−1. A similar graphite walled chamber flushed with CO2 (99.99%) also at a flow rate of 20 ml min−1 served as the neutron insensitive device. Each chamber is calibrated against a national standard 60Co photon source and is operated at an applied voltage of +250 V. Ionization charge is integrated on a Keithley Model 617 electrometer for approximately 60 s. Three successive measurements of the integrated charge were performed at each depth in phantom that typically varied by less than 1% of the average. Measurements were performed with the converter operating at between 75 and 80 kW (CCS fully open and MITR-II operating at 4.5 to 4.8 MW) and have been scaled to the present maximum power in the converter of 83 kW. The response, R, of each chamber in the mixed field normalized by its 60Co calibration factor is described by the following pair of equations (ICRU 1984): Ru = hu D γ + ku D n ˙ ˙ (1) Rt = ht D γ + kt D n . ˙ ˙ (2) The fractional response of each chamber to photons and neutrons of all energies is denoted by h and k, respectively with the subscripts u and t referring to the neutron insensitive and neutron sensitive chambers, respectively. The measured response of each chamber is substituted into ˙ ˙ equations (1) and (2) which are solved for the gamma (D γ ) and neutron (D n ) dose rates when h and k are known for the energy distributions of photons and neutrons in the mixed field. The fast neutron dose rate is then the total neutron absorbed dose rate so obtained less the thermal
  7. 7. Performance characteristics of the MIT FCB 949 Table 1. Summary of the values for ku and kt adopted at each measurement location (in-air and at depth in-phantom). The error specified represents the statistical uncertainty from the MCNP calculation. Measurement position ku kt In-air 0.0102 ± 0.0007 0.786 ± 0.055 10 mm 0.0048 ± 0.0002 0.915 ± 0.049 20 mm 0.0034 ± 0.0002 0.935 ± 0.045 30 mm 0.0024 ± 0.0001 0.942 ± 0.049 40 mm 0.0017 ± 0.0001 0.946 ± 0.052 50 mm 0.0013 ± 0.0001 0.947 ± 0.055 60 mm 0.0010 ± 0.0001 0.948 ± 0.055 neutron dose rate measured by foil activation and scaled by the nitrogen content of A-150 plastic. In the measurements described here values for hu and ht are set to unity as the photon energy spectrum in the mixed field is assumed to be similar to that of the 60Co calibration field. The neutron sensitivities of the graphite and A-150 ionization chambers are given by the following equations (ICRU 1989): ¯ KC WC ku = (3) ¯ Kt WN ¯ KA−150 (sm,g )C WC kt = (4) Kt (sm,g )N W¯N where K represents the kerma coefficient for the chamber materials (in this case graphite and ¯ ¯ A-150 plastic) and Kt is the kerma coefficient for tissue (ICRU 2000). WC and WN denote the average energy expended in creating an ion pair in the gas of the counter in the calibration and the neutron fields, respectively. Values of 29.2 and 32.9 eV have been assumed for WC ¯ in tissue equivalent and CO2 gas, respectively (ICRU 1984). Similarly, (sm,g)C and (sm,g)N represent the mass stopping power ratio for the medium (m) and gas (g) in the calibration and neutron fields, respectively. A stopping power ratio of unity is assumed for both the A-150/TE (methane) and graphite/CO2 chambers (Attix 1986). Values for ku and kt are summarized in table 1 and were determined by averaging the parameters in equations (3) and (4) over the calculated neutron energy spectrum both in-air and at each measuring point in phantom. WN ¯ values for tissue equivalent gas (Jansen et al 1997) and CO2 (Dennis 1973) were weighted by the kerma from the associated energy bin before averaging. A zero slope extrapolation was ¯ applied to the W -values for CO2 below 0.1 MeV, the lowest energy bin calculated by Dennis. 3. Results The calculated and measured in-air epithermal neutron flux (0.5 eV–10 keV) with both the fast neutron and photon absorbed dose rates for each field size are compared in table 2 together with the corresponding specific fast neutron and photon absorbed doses. The epithermal neutron flux varies from 4.6 to 3.2 × 109 n cm−2 s−1 for the aperture sizes currently available that range in diameter between 160 and 80 mm. Beam intensities at the patient position steadily decrease with aperture size as extra snout pieces are added to the end of the collimator, which lengthen the beam line and increase collimation for the smaller fields. The measured specific absorbed doses appear constant for all field sizes with average values of 1.56 and 3.86 × 10−13 Gy cm2 for fast neutrons and photons, respectively, that are well below the background
  8. 8. 950 K J Riley et al Table 2. Summary of measured in-air parameters for each aperture size in the MIT fission converter based epithermal neutron beam. The corresponding calculated values are given beneath each measured result. Aperture Epithermal Fast neutron Photon dose Specific fast Specific photon diameter flux dose rate rate neutron dose dose (mm) (109 n cm−2 s−1) (mGy min−1) (mGy min−1) (10−13 Gy cm2) (10−13 Gy cm2) 160 4.62 ± 0.60 38.2 ± 4.0 96.7 ± 4.2 1.38 ± 0.22 3.49 ± 0.48 4.27 ± 0.10 43.1 ± 2.1 58.7 ± 4.0 1.79 ± 0.09 2.21 ± 0.09 120 4.29 ± 0.56 36.3 ± 3.8 93.7 ± 4.1 1.41 ± 0.22 3.64 ± 0.50 3.83 ± 0.11 38.9 ± 1.3 45.7 ± 0.7 1.69 ± 0.07 1.99 ± 0.06 100 3.41 ± 0.44 34.0 ± 3.5 90.1 ± 3.9 1.66 ± 0.26 4.40 ± 0.60 3.38 ± 0.09 35.8 ± 1.3 41.4 ± 0.8 1.77 ± 0.08 2.04 ± 0.06 80 3.17 ± 0.40 34.1 ± 3.5 74.4 ± 3.2 1.79 ± 0.28 3.91 ± 0.54 2.94 ± 0.08 32.4 ± 1.3 35.2 ± 0.9 1.84 ± 0.09 1.99 ± 0.07 kerma of 2 × 10−12 Gy cm2 inherently produced by epithermal neutrons in tissue. Uncertainties in the calculations include only the statistical error of the tally in MCNP, which is usually smaller than 5%. The measurements have estimated uncertainties expressed as one standard deviation of 15% for epithermal flux, 4.4% for photon and 10.4% for the fast neutron dose components. These uncertainties are less than those stated previously (Rogus et al 1994) due to improved counting statistics and the ability to independently monitor the neutron source strength, namely the converter operating power. The measured neutron flux and dose rates show good agreement with calculations although the measured photon dose rates are sometimes more than double those predicted. This discrepancy is thought to be due to an incomplete description of the photon activity in the model. MCNP does not account for delayed photon emission that occurs for example from 28Al (2.25 min half-life), nor are impurities fully included in the materials of construction (such as antimony, commonly found in lead) the omission of which would reduce the calculated photon absorbed dose rates. Figure 2 shows the vertical cross-plane profiles of the in-air epithermal neutron flux measured at the end of the patient collimator across the 160 mm diameter aperture with the CCS partially raised 0.56 m (converter at 17 kW) and fully open at 1.22 m (converter at 80 kW). This set of measurements was obtained with the MITR-II operating at 4.8 MW. The profiles are each normalized to unity at the maximum observed intensity and have an estimated uncertainty at each point of 1% due to counting statistics and detector positioning. The profiles are symmetric about the central axis of the beam and are nearly rectangular across the aperture. The in-beam intensity falls sharply to approximately 10% of the maximum at a few centimetres outside the boundary of the aperture. The profiles for the FCB operating at 17 and 80 kW are in excellent agreement, demonstrating that the shape of the beam is independent of CCS height and converter power. The full widths at half maximum (FWHM) of these profiles are 170 ± 3 mm which agree well with the 169 ± 4 mm calculated using MCNP. The principal uncertainties in determining the full width half maximum for the calculated profile are statistical uncertainties in the MCNP tallies, while counting statistics and detector positioning dominate the measurement uncertainty. Subsequent gold foil activation and dual ionization chamber measurements performed on the central axis of the beam indicated that the intensities of the individual dose components scaled linearly with the CCS opening and converter power.
  9. 9. Performance characteristics of the MIT FCB 951 Figure 2. Vertical cross-plane profiles of the epithermal neutron flux measured in-air at the collimator aperture for the 160 mm diameter field with a small volume fission counter. The profiles were obtained with the CCS partially and fully opened corresponding to converter powers of 17 and 80 kW (MITR-II at 4.8 MW) respectively. Each profile is normalized to unity at the maximum observed intensity. The measured and calculated depth profiles for the separate dose components in the ellipsoidal water phantom are shown in figure 3 for the four different field sizes. The theoretical predictions are all scaled by 0.83 to normalize the calculated thermal neutron dose to that measured at the depth of the thermal flux maximum for the 160 mm diameter field. Estimates of the experimental uncertainty for the thermal neutron dose component increase from 4.0% near the surface to 6.5% at 100 mm depth in phantom due to poorer counting statistics. The photon component has an uncertainty of 4.4% while that for the fast neutrons varies from 30% at 10 mm depth to 300% at 50 mm. This large uncertainty is due to the inherently small contribution of the fast neutron dose to the total response of the chamber measured in phantom. A statistical uncertainty of 3% is associated with the calculations, which have been omitted from the figure for clarity. The shapes of the calculated curves show generally good agreement with the experimental values although some discrepancy for photons and fast neutrons is apparent at shallow depths near the surface of the phantom. This is most noticeable for the photon absorbed dose rate where the measurements appear uniformly higher than the calculations and is attributed to the incomplete description of the photon production in the beam source. The photon absorbed dose rate is significantly higher in phantom compared to that in-air due to the production of 2.2 MeV photons from the thermal neutron capture reaction with hydrogen that takes place in water. Extrapolating the fast neutron absorbed dose rates measured at depth to the surface of the phantom yields values that are consistent within the experimental uncertainties with those measured in-air. The fast neutron dose rates measured in-air are plotted at a depth of 0 mm in figure 3 for comparison. The photon and thermal neutron dose components exhibit a relatively low entrance dose with a marked build-up occurring at shallow depths that peak between 20 and 30 mm in phantom. Conversely, the fast neutrons have a relatively high entrance dose that rapidly attenuates with depth to such low levels that it is difficult to measure with any reliable experimental uncertainty at depths beyond 40 mm using the dual ionization chamber technique. The total neutron dose rate (thermal plus fast neutrons) measured with the A-150 walled
  10. 10. 952 K J Riley et al Figure 3. Measured (points) and calculated (curves) photon ( ), total neutron ( ), thermal ◦ • neutron ( ) and fast neutron ( ) absorbed dose rates determined in A-150 plastic as a function of depth along the central axis of the ellipsoidal water phantom for each of the four fields (160, 120, 100 and 80 mm diameter). Error bars for the calculations range between 2 and 4% and have been omitted for clarity. All calculated results are scaled by 0.83 to provide a good fit to the measured data. The measured in-air fast neutron dose rates from table 2 are plotted at 0 cm depth ( ) to aid comparison. ionization chamber is similar to that determined from foil activation at depths of 50 mm and greater, indicating that the neutron response of the ionization chamber is essentially due to thermal neutrons at these positions. To describe the performance of the beam as a whole under more realistic conditions similar to clinical irradiations with BPA where boron would be present in tissue the measured dose profiles were weighted with appropriate RBEs to obtain the dose curves shown in figure 4 from which the figures of merit, or the so-called advantage parameters (Clement et al 1990) were determined.
  11. 11. Performance characteristics of the MIT FCB 953 Figure 3. (Continued.) The advantage depth (AD) is defined as the depth where the total weighted dose to the tumour is reduced to the maximum weighted dose received by normal tissue during an irradiation and is a measure of the maximum depth at which therapeutic benefit is obtained. The AD for the 160 mm diameter field is 97 mm and is indicative of a beam possessing good penetration. An AD greater than the average distance of 80 mm to the midline of the head ensures a therapeutic ratio above unity for even the deepest seated brain tumours. An advantage ratio (AR) of 5.7 was obtained from the quotient of the total dose to tumour and that in normal tissue both summed from the surface to the AD. Lastly, the AD dose rate (ADDR) of 1.7 Gy min−1 was determined which is the therapeutic dose rate at the AD and gives the total dose rate achievable to treat tumour at the maximum useful depth of the beam. Design studies estimated an AD, AR and ADDR of 97 mm, 6.2 and 2.0 Gy min−1, respectively (Riley 2001). The small difference between the predicted and actual AR is due to the incomplete description of photon contamination in the MCNP model. The MCNP model also appears to
  12. 12. 954 K J Riley et al Figure 4. RBE weighted dose depth profile measured for the 160 mm diameter field size with the fission converter operating at 83 kW assuming boron concentrations typical for BPA of 65 and 18 µg g−1 for tumour and normal brain tissue, respectively. RBEs of 1.0 for photons and 3.2 for neutrons were applied as well as cRBEs of 1.35 for boron in brain and 3.8 in tumour. overestimate the ADDR by about 17%, which is consistent with the scaling factors applied to fit the calculated results to those measured in figure 3. The lower than expected intensity that is observed is not of practical importance since the present fuel loading in the converter is sufficient to deliver a normal tissue tolerance dose of 12 RBE Gy in just 7 min. To further illustrate the dosimetric properties of the beam figure 5(a) depicts the percentage contribution of the different radiation components to the total RBE weighted normal brain dose as a function of depth when BPA is used as the capture agent. The total photon dose is composed of incident gamma rays and those induced from neutron absorption reactions in tissue. The contribution of the incident component was determined at each depth in phantom using the measured in-air value for photon dose multiplied by attenuation factors calculated with MCNP. Beam contamination accounts for a maximum of 16% of the total weighted dose to normal tissue at 10 mm depth. The remaining 84% of the normal brain tissue dose is induced by events from thermal neutron reactions with boron that has been absorbed during infusion as well as contributions from neutron capture in the constituent nitrogen and hydrogen in normal tissue. The unavoidable photon source from hydrogen capture becomes proportionally more significant with depth and eventually exceeds the boron-related dose at depths greater than approximately 90 mm. The fast neutron dose is small and at 10 mm depth contributes a maximum of only 10% to the total weighted dose, thereafter becoming negligible with depth. Figure 5(b) depicts the relative contribution of each weighted dose component if an advanced porphyrin-based compound is considered. When the FCB is used in conjunction with such a highly selective compound the dose to brain is attributed principally to the inherent kerma produced by neutron induced secondaries following capture reactions with hydrogen and nitrogen nuclei. Beam contamination from incident photons and fast neutrons accounts for no more than 29% of the total weighted dose at 10 mm depth and rapidly decreases with depth to a level comparable to that from boron at about 10% of the total dose. When considering this advanced porphyrin compound, the FCB would achieve an AD of 121 mm, an AR of 10.8 and an ADDR of 0.88 Gy min−1. The influence of the measured beam contaminants on the advantage parameters can be appreciated by setting these components to zero, which would
  13. 13. Performance characteristics of the MIT FCB 955 Figure 5. Percentage contribution of the RBE weighted dose to brain from boron ( ), induced • photons ( ), thermal neutrons ( ), incident photons ( ) and fast neutrons (♦) at each depth in phantom for the fission converter based epithermal neutron beam applicable for (a) BPA and (b) an advanced porphyrin compound. Boron concentrations of 18 and 0.65 µg g−1 are assumed in brain tissue for BPA and the porphyrin, respectively. The applied RBEs are 1.0 for photons and 3.2 for neutrons with cRBEs in brain of 1.35 and 3.8 for BPA and porphyrin, respectively. realize an AD and AR that only modestly improve to 122 mm and 11.5 respectively, while the ADDR would decrease to 0.80 Gy min−1. 4. Discussion A marked improvement in both neutron intensity and beam purity is obtained with the aluminium and Teflon R filtered fission converter based beam compared to that of the previous M67 facility at MIT. The intensity of epithermal neutrons in the new beam has increased by over an order of magnitude from 2 × 108 to 4.6 × 109 n cm−2 s−1. Contamination levels in the FCB are similarly improved and demonstrate the effectiveness of the aluminium
  14. 14. 956 K J Riley et al and Teflon R filter-moderator. Flux levels in the converter are sufficient to produce a high intensity at the patient position after allowing thorough beam filtration to reduce unwanted contaminants. Further reductions in beam contamination are possible, but would not result in practical improvements to the dose profiles achieved in the patient. The FCB is currently the most intense epithermal neutron source for NCT and attains near optimal physical performance in combining high intensity with high beam purity (Harling and Riley 2002, 2003). These attributes should prove highly suitable for not only exploiting the next generation of boron delivery agents but also in providing the possibility of high patient throughput should BNCT become a proved modality. Moreover the facility is currently configured to operate at only a third of its designed and licensed power. This spare capacity could be utilized in the future to maintain the clinically desirable dose rates presently achievable should modifications to the beam line through greater collimation or additional filtration prove advantageous. 5. Summary The performance of the new fission converter based epithermal neutron beam was determined as part of a pre-clinical calibration. The results from a series of dosimetry measurements generally conform to the calculated design specifications. The facility provides variable circular field sizes ranging between 80 and 160 mm in diameter, which in the current configuration produce in-air epithermal fluxes of 3.2 and 4.6 × 109 n cm−2 s−1 respectively at the patient position. The measured specific absorbed doses are constant for all field sizes and are well below the inherent background of 2.8 × 10−12 RBE Gy cm2 produced by epithermal neutrons in tissue. The measured neutron flux and dose rates are in good agreement with calculations although the experimental results for the photon dose rate were sometimes more than double of those predicted which is presumed to be due to inadequacies in the MCNP model. Vertical cross-plane profiles indicate good collimation with a well-defined and symmetric radiation field of uniform intensity, the shape of which is unaffected by the opening height of the CCS. Clinical dose rates can be easily regulated without affecting the character of the beam or varying reactor power using the continuously adjustable CCS in a partially open position. Neutronic beam quality was obtained from depth profiles measured in-phantom for the separate dose components assuming the use of BPA as well as a porphyrin-based boron delivery agent. A therapeutic dose rate of 1.7 Gy min−1 for the 160 mm diameter field is achievable with BPA at the advantage depth of 97 mm for which the average therapeutic ratio is 5.7. If a porphyrin compound is considered, the therapeutic dose rate is reduced to 0.88 Gy min−1 but the advantage depth and advantage ratio (average therapeutic ratio) improve to 121 mm and 10.8, respectively. Neutron capture in the boron nonselectively absorbed in brain following BPA administration will produce the principal dose component affecting normal tissue response that will be observed as an endpoint in the present trials using the FCB. The dose to brain for more selective compounds will predominately result from the inevitable capture of neutrons by the constituent nitrogen and hydrogen. The dose distributions for the aluminium and Teflon R filtered beam approach the theoretical optimum for neutron capture therapy and will be well suited to exploit more advanced boron delivery compounds in the future. The fission converter facility recently constructed at MIT is the first of its kind for BNCT and currently produces the most intense source of epithermal neutrons in the world. High intensity is achieved with high purity to ensure that the dose to normal tissue is principally dependent upon the uptake characteristics of the capture compound being administered. The FCB attains practically ideal beam characteristics that together with its operational flexibility
  15. 15. Performance characteristics of the MIT FCB 957 represent an advance in the state of the art for an epithermal neutron facility. The proved performance characteristics of the FCB coupled with operational versatility make the facility well suited to meet the clinical needs of BNCT in the foreseeable future. Acknowledgments This work has been supported by the US Department of Energy under contract number DEFG02-96ER62193. The authors thank Florent Ghestemme for assisting with calculations of the gold resonance integral. References Ali J 2001 The design, optimization and construction of a patient collimator for the fission converter beam SM Thesis (Cambridge, MA: Massachusetts Institute of Technology) ASTM (American Society for Testing and Materials) 1998 E262–97 Standard Test Method for Determining Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques (West Conshohocken, PA: ASTM) Attix F H 1986 Introduction to Radiological Physics and Radiation Dosimetry (New York: Wiley) Bisceglie E, Colangelo P, Colonna N, Santorelli P and Variale V 2000 On the optimal energy of epithermal neutron beams for BNCT Phys. Med. Biol. 45 49–58 Briesmeister J F 1993 MCNP—A General Monte Carlo N-Particle Transport Code Version 4A (Los Alamos, NM: University of California) Busse P M, Zamenhof R G, Madoc Jones H, Solares G R, Kiger S, Riley K, Chuang C, Rogers G and Harling O K 1997 Clinical follow-up of patients with melanoma of the extremity treated in a phase I boron neutron capture therapy protocol Advances in Neutron Capture Therapy ed B Larsson, J Crawford and R Weinreich (Amsterdam: Elsevier) pp 60–4 Busse P M et al 2002 A critical examination of the results from the Harvard-MIT NCT program Phase I clinical trial of NCT for intracranial disease Research and Development in Neutron Capture Therapy ed W Sauerwein, R Moss and A Wittig (Bologna: Monduzzi) pp 1123–7 Chanana A D et al 1999 Boron neutron capture therapy for glioblastoma multiforma: interim results from phase I/II dose-escalation studies Neurosurgery 44 1182–93 Clement S D, Choi J R, Zamenhof R G, Yanch J C and Harling O K 1990 Monte Carlo methods of neutron beam design for neutron capture therapy at the MIT research reactor (MITR-II) Neutron Beam Design, Development, and Performance for Neutron Capture Therapy ed O K Harling, J A Bernard and R G Zamenhof (New York: Plenum) pp 51–69 Coderre J A, Makar M S, Micca P L, Nawrocky M M, Liu H B, Joel D D, Slatkin D N and Amols H I 1993 Derivations of relative biological effectiveness for the high-LET radiations produced during boron neutron capture irradiations of the 9L rat gliosarcoma in vitro and in vivo Int. J. Radiat. Oncol. Biol. Phys. 27 1121–9 ◦ Dennis J A 1973 Computed ionization and kerma values in neutron irradiated gases Phys. Med. Biol. 18 379–95 Harling O K, Kohse G E and Riley K J 2002a Irradiation performance of polytetrafluoroethylene (Teflon R ) in a mixed fast neutron and gamma radiation field J. Nucl. Mater. 304 83–5 Harling O K and Riley K J 2002 A critical assessment of BNCT beams from fission reactors Research and Development in Neutron Capture Therapy ed W Sauerwein, R Moss and A Wittig (Bologna: Monduzzi) pp 159–62 Harling O K and Riley K J 2003 Fission reactor neutron sources for neutron capture therapy—a critical review J. Neurooncol. at press Harling O K, Roberts K A, Moulin D J and Rogus R D 1995 Head phantoms for neutron capture therapy Med. Phys. 22 579–83 Harling O K et al 2002b The fission converter-based epithermal neutron irradiation facility at the Massachusetts Institute of Technology reactor Nucl. Sci. Eng. 140 223–40 ICRU 1984 Neutron Dosimetry for Biology and Medicine ICRU Report 26 (Bethesda, MD: ICRU) ICRU 1989 Tissue Substitutes in Radiation Dosimetry and Measurement ICRU Report 44 (Bethesda, MD: ICRU) ICRU 2000 Nuclear Data for Neutron and Proton Radiotherapy and for Radiation Protection ICRU Report 63 (Bethesda, MD: ICRU) Jansen J T M, Raaijmakers C P J, Mijnheer B J and Zoetelief J 1997 Relative neutron sensitivity of tissue-equivalent ionisation chambers in an epithermal neutron beam for boron neutron capture therapy Radiat. Prot. Dosim. 70 27–32
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