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PubMed2004

Alanine and TLD coupled detectors for fast neutron dose measurements in neutron capture therapy (NCT).

A method was investigated to measure gamma and fast neutron doses in phantoms exposed to an epithermal neutron beam designed for neutron capture therapy (NCT). The gamma dose component was measured by TLD-300 [CaF2:Tm] and the fast neutron dose, mainly due to elastic scattering with hydrogen nuclei, was measured by alanine dosemeters [CH3CH(NH2)COOH]. The gamma and fast neutron doses deposited in alanine dosemeters are very near to those released in tissue, because of the alanine tissue equivalence. Couples of TLD-300 and alanine dosemeters were irradiated in phantoms positioned in the epithermal column of the Tapiro reactor (ENEA-Casaccia RC). The dosemeter response depends on the linear energy transfer (LET) of radiation, hence the precision and reliability of the fast neutron dose values obtained with the proposed method have been investigated. Results showed that the combination of alanine and TLD detectors is a promising method to separate gamma dose and fast neutron dose in NCT.

Alanine↗
PubMed2002

Controllability of depth dose distribution for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

The updating construction of the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor has been performed from November 1995 to March 1996 mainly for the improvement in neutron capture therapy. On the performance, the neutron irradiation modes with the variable energy spectra from almost pure thermal to epi-thermal neutrons became available by the control of the heavy-water thickness in the spectrum shifter and by the open-and-close of the cadmium and boral thermal neutron filters. The depth distributions of thermal, epi-thermal and fast neutron fluxes were measured by activation method using gold and indium, and the depth distributions of gamma-ray absorbed dose rate were measured using thermo-luminescent dosimeter of beryllium oxide for the several irradiation modes. From these measured data, the controllability of the depth dose distribution using the spectrum shifter and the thermal neutron filters was confirmed.

Antineoplastic Agents↗
PubMed2003

Feasibility study on epithermal neutron field for cyclotron-based boron neutron capture therapy.

To realize the accelerator-based boron neutron capture therapy (BNCT) at the Cyclotron and Radioisotope Center of Tohoku University, the feasibility of a cyclotron-based BNCT was evaluated. This study focuses on optimizing the epithermal neutron field with an energy spectrum and intensity suitable for BNCT for various combinations of neutron-producing reactions and moderator materials. Neutrons emitted at 90 degrees from a thick (stopping-length) Ta target, bombarded by 50 MeV protons of 300 microA beam current, were selected as a neutron source, based on the measurement of angular distributions and neutron energy spectra. As assembly composed of iron, AlF3/Al/6LiF, and lead was chosen as moderators, based on the simulation trials using the MCNPX code. The depth dose distributions in a cylindrical phantom, calculated with the MCNPX code, showed that, within 1 h of therapeutic time, the best moderator assembly, which is 30-cm-thick iron, 39-cm-thick AlF3/Al/6LiF, and 1-cm-thick lead, provides an epithermal neutron flux of 0.7 x 10(9) [n cm(-2) s(-1)]. This results in a tumor dose of 20.9 Gy-eq at a depth of 8 cm in the phantom, which is 6.4 Gy-eq higher than that of the Brookhaven Medical Research Reactor at the equivalent condition of maximum normal tissue tolerance. The beam power of the cyclotron is 15 kW, which is much lower than other accelerator-based BNCT proposals.

Aluminum Compounds↗
PubMed1999

Boron neutron capture therapy (BNCT): implications of neutron beam and boron compound characteristics.

The potential efficacy of boron neutron capture therapy (BNCT) for malignant glioma is a significant function of epithermal-neutron beam biophysical characteristics as well as boron compound biodistribution characteristics. Monte Carlo analyses were performed to evaluate the relative significance of these factors on theoretical tumor control using a standard model. The existing, well-characterized epithermal-neutron sources at the Brookhaven Medical Research Reactor (BMRR), the Petten High Flux Reactor (HFR), and the Finnish Research Reactor (FiR-1) were compared. Results for a realistic accelerator design by the E. O. Lawrence Berkeley National Laboratory (LBL) are also compared. Also the characteristics of the compound p-Boronophenylaline Fructose (BPA-F) and a hypothetical next-generation compound were used in a comparison of the BMRR and a hypothetical improved reactor. All components of dose induced by an external epithermal-neutron beam fall off quite rapidly with depth in tissue. Delivery of dose to greater depths is limited by the healthy-tissue tolerance and a reduction in the hydrogen-recoil and incident gamma dose allow for longer irradiation and greater dose at a depth. Dose at depth can also be increased with a beam that has higher neutron energy (without too high a recoil dose) and a more forward peaked angular distribution. Of the existing facilities, the FiR-1 beam has the better quality (lower hydrogen-recoil and incident gamma dose) and a penetrating neutron spectrum and was found to deliver a higher value of Tumor Control Probability (TCP) than other existing beams at shallow depth. The greater forwardness and penetration of the HFR the FiR-1 at greater depths. The hypothetical reactor and accelerator beams outperform at both shallow and greater depths. In all cases, the hypothetical compound provides a significant improvement in efficacy but it is shown that the full benefit of improved compound is not realized until the neutron beam is fully optimized.

Biophysics↗
PubMed1994

Concomitant boron-neutron capture therapy during fast-neutron irradiation of a rat glioma.

PURPOSE: Fast-neutron irradiation and boron-neutron capture therapy (BNCT) have been independently investigated as treatments for malignant disease. This study tested the feasibility of enhancing fast-neutron irradiation with concomitant BNCT. MATERIALS AND METHODS: Seventeen male Fisher rats, each weighing 180-200 g and bearing 36B10 gliomas, were irradiated with graded doses of fast-neutron radiation. Half of the animals received an L-para-boronophenylalanine (BPA) fructose complex prior to treatment. An in vitro colony-forming assay was used to measure surviving fraction. RESULTS: A significantly lower surviving fraction was noted in the tumors from the BPA group compared with those receiving neutrons alone at the three lower neutron doses (P < .005). With use of a linear quadratic curve fit of cell survival, the dose modifying factor was 1.32 at the 0.10 surviving fraction. Mean tumor boron concentration was 68.4 micrograms/g. CONCLUSIONS: BNCT enhancement of fast-neutron irradiation is feasible in an in vivo tumor system.

Animals↗
PubMed1990

[Neutron flow measurements in the d(14) + Be neutron radiation field from the cyclotron in Essen].

It exists the possibility in the neutron therapy of specific tumours to increase the local energy dose in the tumour by treatment with fast neutrons under use of B-10. Thermal neutrons cause the neutron capture reaction B-10(n, alpha)Li-7 in tissue. A sufficiently high number of thermic neutrons is necessary combined with a possibly selective concentration of B-10 in the tumour as target volume for the clinical application. Gold foils were activated in a water phantom at several depths for the quantitative determination of the thermal neutron fluence. A result of this measurements is a maximal thermal neutron fluence of 1.44 X 10(10) cm-2 at depth of 5 cm with applied total energy dose of 0.8 Gy.

Fast Neutrons↗
PubMed1985

Neutron and neutron boost irradiation of soft tissue sarcomas.

The results of neutron and neutron boost irradiation of 199 patients with soft tissue sarcomas treated between 1978 and 1983 are presented. The median follow-up period is 42 months. The recurrence free survival rates by last review are 93% for patients with T1 tumours (n = 14), 87% for T2 tumours (n = 84) and 73% for T3 tumours (n = 101). The actuarial survival rates at six years are 77% for T1, 63% for T2 and 34% for T3 tumours (p = 0.018). The actuarial survival rate for the group of patients irradiated after surgery without clinical evidence of residual tumour is 63.8% compared with 30.9% for the group of patients with measurable tumour volume at the beginning of radiotherapy (p = 0.002). The survival rates according to grading are 52% for patients with G1 tumours (n = 44), 54% for G2 tumours (n = 130) and 36% for G3 tumours (n = 25). The morbidity rate of 22% after full neutron irradiation was reduced to 15% by the introduction of a neutron boost. At the present time, the results of this modified treatment are not inferior to a full neutron course. The effectiveness of neutron or neutron boost irradiation in the postoperative treatment of soft tissue sarcomas will be evaluated in a forthcoming EORTC trial.

Fast Neutrons↗
PubMed2004

Neutron spectra and angular distributions of concrete-moderated neutron calibration fields at JAERI.

The Facility of Radiation Standards of Japan Atomic Energy Research Institute has been equipped with concrete-moderated neutron calibration fields as simulated workplace neutron fields. The fields use an 241Am-Be neutron source placed in the narrow space surrounded by concrete bricks, walls and floor. The neutron spectra and the neutron fluence rates of the fields were measured with the Bonner multi-sphere spectrometer system (BMS), spherical recoil-proton proportional counters, and a liquid scintillation counter (NE-213). The results were compared with each other. The reference values of H*(10) were determined from the results of BMS. The angular distributions of neutron fluence were calculated using MCNP-4B2 to obtain the reference values of Hp(10). The calculated results show that the scattered neutrons have a wide range of incident angles. The reference Hp(10) values considered the angular distribution were found to be 10-18% smaller than those without consideration.

Algorithms↗
PubMed2005

Preliminary design of a Gd-NCT neutron beam based on compact D-T neutron source.

Gadolinium has been recently proposed, as neutron capture agent in NCT (Neutron Capture Therapy), due to both the nuclide high neutron capture cross section, and the remarkable selective uptake inside tumour tissue that Gd-loaded compounds, can provide. When a neutron external source is supplied, different Gd nuclear reactions, and the generated Auger electrons in particular, cause a high local energy deposition, which results in a tumour cell inactivation. Preliminary micro- as well as macrodosimetric Monte Carlo computational investigations show that the tumour-to-healthy tissue biological damage ratio is in close relation to the neutron beam energy spectrum. The results points out that the optimum neutron spectrum, to be used for Gd-NCT, seems to lie in the 1 to 10 keV energy range. In order to 'tailor' such spectra, an original, accelerator-driven, neutron source and spectrum shaping assembly for hospital-based Gd-NCT are presented and preliminary results are reported.

Brain Neoplasms↗
PubMed1984

Environmental neutron measurements around nuclear facilities with moderated-type neutron detector.

A cylindrical BF3 counter with cylindrical polyethylene moderators of 1-, 3-, 6- and 10-cm wall thickness were developed for neutron spectrometry, having response functions calculated by the time-dependent multi-group Monte Carlo code, TMMCR, and evaluated by a time-of-flight experiment in the energy region from epithermal to about 50 keV. This system of neutron detectors was used to measure the environmental neutron spectrum and dose equivalent around a 252Cf fission neutron source, a fast neutron source reactor, a cyclotron and an electron synchrotron. The neutron spectrum and dose obtained by this detector system showed very good agreement with those measured by a commercially available neutron rem counter and calculated by our MMCR Monte Carlo code. As a result of these comparisons, this detector system has been established as a standard spectrometer and dosimeter for environmental measurements around nuclear facilities.

Air Pollution, Radioactive↗
PubMed1994

Benchmark test of neutron transport calculations: indium, nickel, gold, europium, and cobalt activation with and without energy moderated fission neutrons by iron simulating the Hiroshima atomic bomb casing.

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a bare- and energy-moderated 252Cf fission neutron source which was obtained by transmission through 10-cm-thick iron. An iron plate was used to simulate the effect of the Hiroshima atomic bomb casing. This test includes the activation of indium and nickel for fast neutrons and gold, europium, and cobalt for thermal and epithermal neutrons, which were inserted in the moderators. The latter two activations are also to validate 152Eu and 60Co activity data obtained from the atomic bomb-exposed specimens collected at Hiroshima and Nagasaki, Japan. The neutron moderators used were Lucite and Nylon 6 and the total thickness of each moderator was 60 cm or 65 cm. Measured activity data (reaction yield) of the neutron-irradiated detectors in these moderators decreased to about 1/1,000th or 1/10,000th, which corresponds to about 1,500 m ground distance from the hypocenter in Hiroshima. For all of the indium, nickel, and gold activity data, the measured and calculated values agreed within 25%, and the corresponding values for europium and cobalt were within 40%. From this study, the MCNP code was found to be accurate enough for the bare- and energy-moderated 252Cf neutron activation calculations of these elements using moderators containing hydrogen, carbon, nitrogen, and oxygen.

Cobalt↗
PubMedDate not supplied

Optimization of boron and neutron delivery for neutron capture therapy.

A number of groups in the United States have received funding that will permit evaluation of the clinical efficacy of the neutron capture therapy (NCT) procedure. Various reactors are being modified to allow the construction of an epithermal neutron beam. At the Brookhaven Medical Research Reactor (BMRR), the patient irradiation facility is being modified to produce an optimized epithermal neutron beam. An 80-cm-thick A1-D2O mixture (184 g/cm2, 25% D2O by volume) is being installed in the shutter assembly. One-dimensional calculations indicate that this configuration should provide an epithermal neutron flux density of approximately 1 x 10(9) n/cm2/sec at 3 MW and a concomitant fast neutron dose rate of approximately 2 x 10(-11) rad per epithermal neutron (assuming a homogeneous A1-D2O mixture). The actual geometry will be an inhomogeneous array of D2O and A1 layers producing parameters somewhat less favorable than those listed above; experimental verification is in progress. Significant gains have recently been made in selectively targeting B to melanoma with various melanaffinic compounds, including p-boronophenylalanine, and with boronated porphyrins that may be applicable to a variety of tumors. Neutron capture radiographs have been obtained with the above compounds, and efforts have been made to quantitate boron uptake in growing and quiescent or necrotic regions of tumor via double-labeling techniques obtained with tritiated thymidine. A correlation between therapeutic efficacy and the ability to deliver boron to viable areas of tumor has been observed.

Animals↗
PubMedDate not supplied

A prototype epithermal neutron beam for boron neutron capture therapy.

An epithermal neutron beam has been designed and tested at the Georgia Institute of Technology's 5-MW Research Reactor. The prototype facility consists of aluminum and sulfur disks in a tangential beam port for fast neutron filtration. A cadmium sheet at the port exit removes the thermal neutrons from the transmitted beam, leaving an intensely epithermal neutron beam spanning five energy decades, each contributing to the flux demanded by boron neutron capture therapy. The thermal neutron flux generated by the incident epithermal neutrons in a polyethylene head phantom peaks at a depth of 3 cm and remains above the incident thermal flux to a 7-cm depth. The beam thus provides the penetration required for treating deep-seated gliomas. Photon contamination in the prototype facility is high, and a number of basic modifications are proposed for reducing it to safer levels.

Boron↗
PubMedDate not supplied

Accelerator-based epithermal neutron beam design for neutron capture therapy.

Recent interest in the production of epithermal neutrons for use in boron neutron capture therapy (BNCT) has promoted an investigation into the feasibility of generating such neutrons with a high current proton accelerator. Energetic protons (2.5 MeV) on a 7Li target produce a spectrum of neutrons with maximum energy of roughly 800 keV. A number of combinations of D2O moderator, lead reflector, 6Li thermal neutron filtration, and D2O/6Li shielding will result in a useful epithermal flux of 1.6 x 10(8) n/s at the patient position. The neutron beam is capable of delivering 3000 RBE-cGy to a tumor at a depth of 7.5 cm in a total treatment time of 60-93 min (depending on RBE values used and based on a 24-cm diameter x 19-cm length D2O moderator). Treatment of deeper tumors with therapeutic advantage would also be possible. Maximum advantage depths (RBE weighted) of 8.2-9.2 (again depending on RBE values and precise moderator configuration) are obtained in a right-circular cylindrical phantom composed of brain-equivalent material with an advantage ratio of 4.7-6.3. A tandem cascade accelerator (TCA), designed and constructed at Science Research Laboratory (SRL) in Somerville MA, can provide the required proton beam parameters for BNCT of deep-seated tumors. An optimized configuration of materials required to shift the accelerator neutron spectrum down to therapeutically useful energies has been designed using Monte Carlo simulation in the Whitaker College Biomedical Imaging and Computation Laboratory at MIT. Actual construction of the moderator/reflector assembly is currently underway.

Boron↗
PubMed1995

Determination of the thermal neutron flux in a fast neutron beam by use of a boron-coated ionization chamber.

The thermal neutron distribution in slow and fast neutron beams is usually determined using the foil activation method. In this work a small magnesium walled ionization chamber, in which the inner surface of the wall has been coated with 10B to increase the sensitivity for thermal neutrons, is used to estimate the thermal neutron component of the beam. After calibration and determination of the directional response in a thermal neutron beam a comparison with foil activation at different depths in water was performed to investigate the reliability of the ionization measurements. The chamber was used in a computer controlled water phantom to measure the depth and lateral distribution of the thermal neutron dose. With this arrangement two-dimensional scans of the thermal neutrons could be performed quickly and with high accuracy.

Biophysical Phenomena↗
PubMed1999

Paired Mg and Mg(B) ionization chambers for the measurement of boron neutron capture dose in neutron beams.

The use of the boron neutron capture (BNC) reaction to provide a dose enhancement in fast neutron therapy is currently under investigation at the Gershenson Radiation Oncology Center of Harper Hospital in Detroit, MI. The implementation of this treatment modality presents unique challenges in dosimetry. In addition to the measurement of photon and neutron doses in the mixed field, a measure of the thermal neutron flux and the associated boron neutron capture dose throughout the treatment volume is desired. A pair of small-volume magnesium ionization chambers has been constructed with the aim of providing this information. One of the chambers, denoted the Mg(B) chamber, is lined with a boron-loaded foil. The ionization response of this chamber has been calibrated in terms of BNC dose per ppm loading of 10B. These paired chambers can be used to map the local BNC response in neutron beams. From this data and an estimation of the boron concentration in the tumor and normal tissue, the boron neutron capture enhancement may be evaluated.

Boron↗
PubMed2001

Determination of radionuclides induced by fast neutrons from the JCO criticality accident in Tokai-mura, Japan for estimating neutron doses.

A criticality accident occurred at a uranium conversion facility in Tokai-mura, Japan on September 30, 1999, and fission neutrons were continuously emitted for about 20 hours. Materials of stainless steel or iron, and chemical reagents were collected at places between 2 m and 270 m from the criticality accident site on October 25 and 26, 1999, November 27, 1999 and February 11, 2000. Neutron-induced radionuclides. such as 54Mn and 58Co, in the materials exposed to fast neutrons from the accident were measured to estimate the neutron fluences and energy distributions. Highly sensitive y-ray spectrometry with a well-type Ge detector was performed after radiochemical separation of Mn and Co from the materials. An instrumental neutron activation analysis was mainly applied for determinations of the target elements and chemical yields. The concentrations of 54Mn and 58Co in a mesh screen of stainless steel collected at a location 2.0 m from the accident site were determined. The total number of fission events was evaluated to be 2.5 x 10(18) by Monte-Carlo calculations of neutron transfer by considering the observed values of 54Mn and 58Co. The results presented here are fundamental to estimate the neutron doses at various distances.

Fast Neutrons↗
PubMed1986

[Shielding design and detection of neutrons from medical and industrial electron accelerators--simple method of design calculation for neutron shielding].

The neutron leakage from medical and industrial electron accelerators has become an important problem and its detection and shielding is being performed in their facilities. This study provides a new simple method of design calculation for neutron shielding of those electron accelerator facilities by dividing into the following five categories; neutron dose distribution in the accelerator room, neutron attenuation through the wall and the door in the accelerator room, neutron and secondary photon dose distributions in the maze, neutron and secondary photon attenuation through the door at the end of the maze, neutron leakage outside the facility-skyshine.

Electrons↗