Mixed Neutron Research Articles

Mixed photon and neutron radiotherapy given concurrently with chemotherapy in unresectable pancreatic cancer.

328 Background: Unresectable tumors of the pancreas remain difficult to treat despite the advent of targeted radiotherapy and modern chemotherapy. Randomized trials exploring the efficacy of chemotherapy and radiation have demonstrated median survival of 9 to 11 months. These survival times have not improved appreciably in the modern era. The purpose of this study was to retrospectively review our institutional experience with unresectable pancreatic cancer treated with mixed photon-neutron radiotherapy given concurrently with chemotherapy. Methods: Thirteen patients with unresectable tumors of the pancreas were treated between 1993 and 2001. All patients were treated with mixed photon-neutron radiotherapy given concurrently with chemotherapy. Median photon dose was 39.6 Gy (30.6-45Gy) and median neutron dose was 8 nGy (7-9 nGy). 12 of 13 patients were treated with neoadjuvant chemotherapy, followed by 5-FU given concurrently with radiotherapy. Median survival, overall survival, and local control were calculated for all patients. Results: The median age of all patients was 65 years (46-75 years). Twelve patients had histologic diagnosis of adenocarcinoma, with the other having an islet cell carcinoma. All patients are now deceased. Median survival for all patients was 11.5 months (3.0-25.6 months). The 1 and 2- year overall survival was 46.2% and 7.7%, respectively. Local control of the primary tumor was excellent at 92.3%. The rate of distant metastasis was 76.9%. One patient experienced decline without documented recurrence. No grade ≥3 acute toxicities were reported. However, there were 2 grade 5 late toxicities, both caused by gastrointestinal bleeding. Conclusions: Our experience demonstrates that treatment of unresectable pancreatic tumors with mixed photon-neutron radiotherapy given concurrently with chemotherapy results in excellent local control, with survival time equivalent to or exceeding that demonstrated in previous series. With the added capability of intensity modulated neutron radiation therapy (IMNRT), the incidence of treatment-related morbidity may be improved while taking advantage of the superior tumor control that high-LET radiation may impart. No significant financial relationships to disclose.

Carrier-Removal Comparison (n/p) and Functional Testing of Si and SiC Power Diodes

Commercial Si and 4H-SiC Schottky barrier power diodes were irradiated in the mixed neutron and gamma-ray radiation field of The Ohio State University research reactor (OSURR). The forward I-V characteristics were measured before and immediately after each successive irradiation, and the carrier-removal rates were compared, on the basis of NIEL, to a previous study, for which the same diode models were irradiated with a 203 MeV proton beam. In addition, a number of SiC Schottky barrier diodes were also irradiated in the OSURR and subsequently functionally tested in half-wave rectifier circuits, for which the voltage and current waveforms in the circuit were recorded. The results from the functional testing of these half-wave rectifier circuits were analyzed using results from I-V characterization, PSpice simulations, and an analytical formulation.

Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz

To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the 7Li(n,α)3H-reaction. These particles are detected by the diode and their amount correlates to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC-calculations for mixed radiation fields and the Hansen & Olsen alanine detector response model. With the acquired data about the background dose and charged particle spectrum, and with the acquired information of the neutron flux, we are capable of calculating the dose to the tissue. Conclusion. Monte Carlo simulation of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also speculate on sensitizing alanine to thermal neutrons by adding boron compounds.

Study of the effect of high dose rate on tissue equivalent proportional counter microdosimetric measurements in mixed photon and neutron fields

Abstract This study describes the measurement of lineal energy spectra carried out with a 5.1 cm (2 in.) diameter spherical tissue equivalent proportional counter (TEPC) simulating 2 μm tissue equivalent (TE) site diameter in low energy mixed photon–neutron fields with varying dose rates generated by employing the McMaster University 1.25 MV double stage Tandetron accelerator. The 7 Li (p, n) 7 Be reaction was employed to generate a variety of mixed fields of photons and low energy neutrons using proton beam energy ranging 1.89–2.56 MeV. The dose rate at a given beam energy was varied by changing the beam current. Dose rates that resulted in dead times as high as 75% were employed to study the effect of dose rate on quality, microdosimetric averages ( y¯ F and y¯ D ), absorbed dose and dose equivalent. We have observed that high dose rates due to both photons and neutrons in a mixed field of radiation result in pile up of pulses and distort the lineal energy spectrum measured under these conditions. The pile up effect and hence the distortion in the lineal energy spectrum becomes prominent with dose rates, which result in dead times larger than 25% for the high linear energy transfer (LET) radiation component. Intense neutron fields, which may amount to 75% dead time, could result in a 50% or even larger increase in the values of the microsdosimetric averages and the neutron quality factor. This study demonstrates moderate dose rates that do not result in dead times of more than 20–25% due to either of the component radiation or due to both components of mixed field radiation generate results that are acceptable for radiation monitoring.

Spatial frequency analysis of fluorescent nuclear track detectors irradiated in mixed neutron–photon fields

Abstract A new image processing method for the dosimetry of mixed neutron–photon fields using fluorescent nuclear track detectors (FNTDs) is investigated. Images obtained from FNTDs were processed using spatial frequency analysis to determine doses for both low and high LET radiations. Spatial frequency analysis extended the dynamic range of detectible neutron doses from 4 orders of magnitude for track counting only to at least 6 orders of magnitude by combining track counting with the new image processing method. Two different converters of indirectly ionizing radiation to secondary charged particles were used in conjunction with image processing to separate signals induced by neutron and gamma fields.

NXcom – A program for calculating attenuation coefficients of fast neutrons and gamma-rays

This work is concerned with a construction and use of NXcom computer program for calculating the removal and attenuation coefficients of transmitted fast neutrons and γ-rays, respectively, through mixtures, composites, concretes and compounds. The program uses only one input data file for neutrons and γ-rays calculations. For γ-ray attenuation, the program predictions were tested by comparing them with the well-known WinXcom program results and an excellent agreement was noticed. Also, it has been used for calculating the values of macroscopic effective removal cross-sections Σ R (cm −1) for five new published polyamide and anhydride composites designed for shielding mixed neutron and γ-rays. The obtained values for Σ R using the program and the reported attenuation thicknesses which were based on the Monte Carlo N-Particle (MCNP) code showed the same trend. The NXcom program can be used as a preliminary effective tool for testing the shielding material against fast neutrons and γ-rays.

Publisher’s Note: Using theΔ3statistic to test for missed levels in mixed sequence neutron resonance data [Phys. Rev. C80, 034612 (2009)

Publisher’s Note: Using the<mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" display="inline"><mml:mrow><mml:msub><mml:mi>Δ</mml:mi><mml:mrow><mml:mn>3</mml:mn></mml:mrow></mml:msub></mml:mrow></mml:math>statistic to test for missed levels in mixed sequence neutron resonance data [Phys. Rev. C<b>80</b>, 034612 (2009)

A discrete ordinates operator for the diffuse transmission of a mixed neutron beam through a multilayered medium

Abstract In this article the transport model and inductive approach reported in a recent work have been extended, and a discrete ordinates operator for the diffuse transmission of a mixed neutron beam through a multilayered medium is derived. At first, a mathematical description of the mixed beam transport problem that we are concerned with is given. The resulting mathematical problem is, according to Chandrasekhar, decomposed into uncollided and diffuse problems. After treating the uncollided problem analytically, a discrete ordinates formulation of the diffuse problem is considered. A set of spatially discretised equations that are free from spatial truncation error is then introduced. Finally, a discrete ordinates operator that replaces the entire multilayered medium in diffuse neutron transmission computations is derived inductively. To the best of our knowledge, such an operator has never been reported previously in neutron transport literature. Numerical results for a four-layer medium are presented to illustrate the computational merit of a computer code based on our diffuse transmission operator.

Sample characterization using both neutron and gamma multiplicities

Formulas for multiplicity counting rates (singles, doubles, etc.), used for the unfolding of parameters of an unknown sample, can be derived from those for the corresponding factorial moments. So far such rates were derived only for neutrons. The novelty of this paper is related to the derivation of the individual gamma and mixed neutron–gamma detection rates as well as to investigation of the possibilities and actual algorithms for the sample parameter unfolding. Taking the individual gamma and mixed neutron–gamma moments up to third order, together with the neutron moments, there will be nine auto- and cross-factorial moments and corresponding multiplicity rates, as well as a larger number of unknowns than for pure neutron detection. The total number of measurable multiplicities exceeds the number of unknowns, but on the other hand the structure of the additional equations is substantially more complicated than that of the neutron moments. Since an analytical inversion of the highly non-linear system of over-determined equations is not possible, the use of artificial neural networks (ANNs) is suggested, which can handle both the non-linearity and the redundance in the measured quantities in an effective and accurate way. The use of ANNs is demonstrated with good results on the unfolding of various combination of multiplicities for certain combinations of the unknown parameters, including the sample fission rate, the leakage multiplication and the α and γ ratios.

Calculation of responses and analysis of experimental data for a silicon gamma spectrometer

A silicon detector developed for gamma ray spectrometry (intended for measurement in the mixed neutron gamma fields, e.g. in reactor physics) was tested for the photon energy interval up to 10 MeV. The responses of this spectrometric system were simulated using a Monte Carlo method (MCNPX code), and the response matrix was calculated. Unfolding techniques for reconstructing incident photon spectra were tested, and a method based on the Scofield–Gold iterative solution of the corresponding matrix equation was applied. The applicability of such an approach to spectrometry data processing and spectral analysis is demonstrated, an example of the results is presented, and possible future improvements and optimizations are discussed.

On Computing Neutron Transmission from the Responses of a Void Fraction Evaluation Scheme with Mixed Beams

In this technical note we report on a slight but important modification in a recently developed backscattered neutron-based void fraction evaluation scheme for slab materials, and we describe an add-on numerical scheme for computing total (direct plus diffuse) neutron transmission through a test slab. In the void fraction evaluation scheme, the broad neutron beam consists of a monodirectional (singular), normally incident component and a smooth (regular), angularly continuous component, i.e., a mixed neutron beam. Once the void fraction of the test slab has been evaluated, the diffuse component of the angular flux of transmitted neutrons can be computed from an accurate spherical harmonics–discrete ordinates solution of the neutron beam transport problem defined in a reduced slab domain (the direct component is rather straightforward to compute). The add-on scheme described here can be used to evaluate the amount of neutrons that escape from the slab through the back side. Numerical results are given to illustrate the usefulness of our add-on scheme in neutron shielding studies.

Test beam results of heavily irradiated magnetic Czochralski silicon (MCz-Si) strip detectors

Strip detectors with an area of 16 cm 2 were processed on high resistivity n-type magnetic Czochralski silicon. In addition, detectors were processed on high resistivity Float Zone wafers with the same mask set for comparison. The detectors were irradiated to several different fluences up to the fluence of 3 × 10 15 1 MeV n eq / cm 2 with protons or with mixed protons and neutrons. The detectors were fully characterized with CV- and IV-measurements prior to and after the irradiation. The beam test was carried out at the CERN H2 beam line using a silicon beam telescope that determines the tracks of the incoming particles and hence provides a reference measurement for the detector characterization. The n-type MCz-Si strip detectors have an acceptable S/N at least up to the fluence of 1 × 10 15 n eq / cm 2 and thus, they are a feasible option for the strip detector layers in the SLHC tracking systems.

Monte Carlo estimation of photoneutrons contamination from high-energy X-ray medical accelerators in treatment room and maze: a simplified model

Despite all advantages associated with high-energy radiotherapy to improve therapeutic gain, the production of photoneutron via interaction of high-energy photons with high atomic number (Z) materials increases undesired dose to the patient and staff. Owing to the limitation and complication of experimental neutron dosimetry in mixed beam environment, including photon and neutron, the Monte Carlo (MC) simulation is a gold standard method for calculation of photoneutron contaminations. On the other hand, the complexity of treatment head makes the MC simulation more difficult and time-consuming. In this study, the possibility of using a simplified MC model for the simulation of treatment head has been investigated using MCNP4C general purpose MC code. As a part of comparative assessment strategy, the fluence, average energy and dose equivalent of photoneutrons were estimated and compared with other studies for several fields and energies at different points in treatment room and maze. The mean energy of photoneutrons was 0.17, 0.19 and 0.2 MeV at the patient plan for 10, 15 and 18 MeV, respectively. The calculated values differed, respectively, by a factor of 1.4, 0.7 and 0.61 compared with the reported measured data for 10, 15 and 18 MeV. Our simulation results in the maze showed that the neutron dose equivalent is attenuated by a factor of 10 for every 4.6 m of maze length while the related factor from Kersey analytical method is 5 m. The neutron dose equivalent was 4.1 mSv Gy(-1) at the isocentre and decreased to 0.79 mSv Gy(-1) at a distance of 100 cm away from the isocentre for 40 x 40 cm(2). There is good agreement between the data calculated using simplified model in this study and measurements. Considering the reported high uncertainties (up to 50%) in experimental neutron dosimetry, it can be concluded that the simplified model can be used as a useful tool for estimation of photoneutron contamination associated with high-energy photon radiotherapy.

Material void fraction evaluation with mixed neutron beams: theory and numerical experiment

The aim of this article is to describe a mathematical model and an associated numerical method for the basic problem of evaluating the void fraction of a slab material from the responses of a typical source-detector system for backscattered neutrons. The mathematical model and void fraction evaluation method reported herein: i) enable us to treat explicitly and exactly a mixed neutron beam, i.e. a neutron beam consisting of a monodirectional component with normal incidence upon the slab material and a component that is a smooth function of the angular variable and ii) allow for an arbitrary order of anisotropy of the neutron scattering phase function. In addition, the associated void fraction evaluation method is free from computer overflow exceptions, even for the challenging problems where the angular quadrature set is of high order and/or the slab material is exceedingly thick. We perform a numerical experiment to illustrate the positive features of both the mathematical model and evaluation method reported here, and we conclude this article with a discussion and directions for future work.

BNCT for skin melanoma in extremities: Updated Argentine clinical results

As part of phase I/II melanoma BNCT clinical trial conducted in Argentina in a cooperative effort of the Argentine Atomic Energy Commission (CNEA) and the Oncology Institute Angel H. Roffo (IOAHR), 7 patients (6 female–1 male) received eight treatment sessions covering ten anatomical areas located in extremities. Mean age of the patients was 64 years (51–74). The treatments were performed between October 2003 and June 2007. All patients presented multiple subcutaneous skin metastases of melanoma and received an infusion containing ∼14 gr/m 2 of 10borophenyl-alanine (BPA) followed by the exposition of the area to a mixed thermal-epithermal neutron beam at the RA-6 reactor. The maximum prescribed dose to normal skin ranged from 16.5 to 24 Gy-Eq and normal tissue administered dose varied from 15.8 to 27.5 Gy-Eq. Considering evaluable nodules, 69.3% of overall response and 30.7% of no changes were seen. The toxicity was acceptable, with 3 out of 10 evaluable areas showing ulceration (30% toxicity grade 3).

On the readout of HT peaks in LiF:Mg,Ti by applying high rate gas heating

The main high temperature (HT) peaks in the glow curves of LiF:Mg,Ti, normally assigned as 6 and 7 are known to have higher response to high LET radiation than to low LET radiation. This effect can be used in principle to discriminate between neutrons and gamma rays in a mixed neutron–gamma field. In routine dosimetry, a high rate gas heating of Teflon coated chips is employed (mostly 25 °C/s), which may cause marginal thermal conditions. The fast heating rate readout process of the HT peaks and the practical consequences were studied, leading to the conclusion that in spite of the technical constraints, using high rate gas heating produces results which are not significantly different from those using a slow heating rate. Applying high heating rates seems to pose no practical obstacle in utilizing HT peaks information, which has the potential to be used for mixed neutron–gamma dosimetry.

Radiation effects in polycarbonate capacitors

The aim of this paper is to examine the influence of neutron and gamma irradiation on the dissipation factor and capacitance of capacitors with polycarbonate dielectrics. The operation of capacitors subject to extreme conditions, such as the presence of ionizing radiation fields, is of special concern in military industry and space technology. Results obtained show that the exposure to a mixed neutron and gamma radiation field causes a decrease of capacitance, while the loss tangent remains unchanged.

Helium and hydrogen measurements on pure materials irradiated in SINQ Target 4

Several irradiations have been performed in the Swiss Spallation Neutron Source (SINQ) to establish a materials database for mixed proton and neutron fluxes for future spallation neutron and other accelerator sources. Pure metal dosimetry materials from the second irradiation (STIP-II) have been analyzed for their total helium and hydrogen contents and their release characteristics with temperature (TDS). Total helium results are similar to those observed earlier from the first irradiation experiment (STIP-I), with concentrations ranging from ∼500 to ∼1000 appm. Hydrogen contents varied over a larger range from ∼100 to ∼60 000. 3He/ 4He ratios were generally consistent with expectations, except for Ti, Nb, and Ta which showed lower values due to 3He from decay of irradiation-generated tritium. Some differences were observed in the hydrogen TDS data for the control and irradiated materials, including some evidence for additional lower-temperature release and for multiple release peaks. Additionally, differences were noted in the releases for irradiated material that been cleaned versus material that had no cleaning.

Dose–response relationship of dicentric chromosomes in human lymphocytes obtained for the fission neutron therapy facility MEDAPP at the research reactor FRM II

The biological effectiveness of neutrons from the neutron therapy facility MEDAPP (mean neutron energy 1.9 MeV) at the new research reactor FRM II at Garching, Germany, has been analyzed, at different depths in a polyethylene phantom. Whole blood samples were exposed to the MEDAPP beam in special irradiation chambers to total doses of 0.14-3.52 Gy at 2-cm depth, and 0.18-3.04 Gy at 6-cm depth of the phantom. The neutron and gamma-ray absorbed dose rates were measured to be 0.55 Gy min(-1) and 0.27 Gy min(-1) at 2-cm depth, while they were 0.28 and 0.25 Gy min(-1) at 6-cm depth. Although the irradiation conditions at the MEDAPP beam and the RENT beam of the former FRM I research reactor were not identical, neutrons from both facilities gave a similar linear-quadratic dose-response relationship for dicentric chromosomes at a depth of 2 cm. Different dose-response curves for dicentrics were obtained for the MEDAPP beam at 2 and 6 cm depth, suggesting a significantly lower biological effectiveness of the radiation with increasing depth. No obvious differences in the dose-response curves for dicentric chromosomes estimated under interactive or additive prediction between neutrons or gamma-rays and the experimentally obtained dose-response curves could be determined. Relative to (60)Co gamma-rays, the values for the relative biological effectiveness at the MEDAPP beam decrease from 5.9 at 0.14 Gy to 1.6 at 3.52 Gy at 2-cm depth, and from 4.1 at 0.18 Gy to 1.5 at 3.04 Gy at 6-cm depth. Using the best possible conditions of consistency, i.e., using blood samples from the same donor and the same measurement techniques for about two decades, avoiding the inter-individual variations in sensitivity or the differences in methodology usually associated with inter-laboratory comparisons, a linear-quadratic dose-response relationship for the mixed neutron and gamma-ray MEDAPP field as well as for its fission neutron part was obtained. Therefore, the debate on whether the fission-neutron induced yield of dicentric chromosomes increases linearly with dose remains open.

Study on Composite Material for Shielding Mixed Neutron and $\gamma$-Rays

Optimized lightweight, compact and high temperature sustaining shielding materials for neutron and gamma radiation were developed by genetic algorithms (GA) combined with the Monte Carlo N-Particle (MCNP) code. A series of samples were designed according to the method. Deep penetration tests by the MCNP code were completed. The results show that the designed samples have more advantages related to the radiation shielding effects in comparison with PB202 and KRAFTON-XP3, especially the Cakes with multi-layers structures of the Fe-interlayer-Pb have excellent performance. Taking into account the ratios designed of components among elements in the material, the manufacturing process of polymer with nano-TiO2 was studied experimentally by differential scanning calorimetry (DSC). Several samples have been tested. The attenuation experiments on the samples were carried out using tandem electrostatic accelerator neutron source, spontaneous fission neutron source of 252Cf and gamma-rays source of 60Co. The experimental results verify the correctness of optimal design and craft. The Cakes are quite suitable for applications in the practices of nuclear science and technology.