Neutron Angular Distributions Research Articles

Dynamics of nuclear interaction in the reaction 56Fe(α, xn)

The spectra and angular distributions of neutrons from the reaction 56Fe(α, xn) were measured at the alpha-particle energies of 12.3, 16.3, 18.3, 26.8, and 45.2 MeV. The measurements were performed with time-of-flight fast-neutron spectrometers at pulsed accelerators of charged particles. The measured data were analyzed within models of equilibrium, preequilibrium, and direct reaction mechanisms. The calculations were performed by using an exact formalism of Hauser-Feshbach statistical theory. The required nuclear-level densities in nickel isotopes excited in this reaction were determined on the basis of new experimental data on low-lying levels, neutron resonances, and spectra of evaporated neutrons. The contributions of equilibrium, preequilibrium, and direct neutron emission were studied over a broad range of alpha-particle energies.

SU-FF-T-624: Skeletal Neutron Dose Response Function Development for Use in Proton Therapy

Purpose: To develop a set of human skeletal neutron absorbed dose response functions for an inhomogeneous, multi-region bone model and neutron energies ranging from 10 eV to 150 MeV for use in neutron dose assessment in proton therapy. Method and Materials: The neutron absorbed dose response function equation was formulated after reviewing ICRU Report 63 and a previously-published photon absorbed dose response function formulation. The neutron absorbed dose response function equation is a function of the neutron energy, the isotope with which the neutron interacts, the emission particle considered, and the source and target bone regions. The first reaction considered was neutron scattering on hydrogen. The neutron angular distribution and elastic scatter cross-section were retrieved from the Evaluated Nuclear Data File. After optimization for numerical integration, proton specific absorbed fraction data was used to calculate the neutron absorbed dose response function for 13 skeletal sites. The neutron kerma response function was also calculated for comparison. This procedure will be extended to neutron-induced proton production in the remaining elements present in the human skeleton. Results: The relative error of the neutron kerma response function with respect to the neutron dose response function is heavily dependent upon skeletal site. It ranges from 0.07 for the sacrum to 0.52 for the femur and the humerus. The hydrogen component of the neutron dose response function is small at low neutron energies due to thermal neutron proton production in nitrogen, comprises nearly all of the neutron dose response function at medium neutron energies, and decreases for high neutron energies. Conclusion: The neutron kerma dose response function is a good approximation for the neutron dose response function at low neutron energies, as the emitted protons do not travel far before they deposit all of their energy, and at high energy, as charged particle equilibrium is approached.

SU‐FF‐T‐277: Measuring High Energy Neutrons at a 230 MeV Proton Therapy Facility

Purpose: Examine neutron field spectrum and angular distribution characteristic to medium‐energy (230 MeV) proton accelerator facilities. Measure effectively the high energy neutron component of the radiation field. Method and Materials: Primary shielding walls in modern proton therapy facilities are typically made of an over 7′ thick of ordinary (2.3 g/cm3) concrete. Our analysis demonstrate that behind such a thick wall in forward direction with respect to the proton beam neutrons with energies greater than 8 MeV still contribute considerably to the total dose. High energy neutron contribution in lateral direction is potentially even larger. Traditionally neutron rem‐meters are designed to have their response function match well an appropriate (ICRP 1990, or NCRP‐38) fluence‐to‐dose conversion function over an energy range extending from thermal (0.025 ev) to 10 MeV. We have performed over thirty measurements of the neutron dose equivalent at various locations in and around a state‐of‐the‐art 230 MeV proton therapy facility using the Wide Energy Neutron Detection Instrument (WENDI) that has a useful energy response in the energy range from thermal to 5 GeV. Results: We have obtained neutron attenuation lengths in forward and lateral directions from our measurements most appropriate for use at 230 MeV proton therapy facilities (PTFs) and compared our results with some previously published values. We have also obtained a new parameterization for neutron attenuation in the maze suitable for modern PTFs. Conclusion: In surveying and area monitoring modern PTFs the neutron detector of choice must be capable of detecting with sufficient efficiency the high energy component of the neutron field to avoid large dose underestimation in an environment where greater than 8 MeV neutrons may have a significant contribution. New maze attenuation parameterization will provide for adequate maze design in these facilities, where previously used attenuation models provided at best for a marginal maze design.

Optimized therapeutic neutron beam for accelerator-based BNCT by analyzing the neutron angular distribution from 7Li(p,n) 7Be reaction

Perpendicular neutrons (i.e., solid angle bin of 50–150°) among ones generated from 7Li(p,n) 7Be reaction were used to produce an optimized therapeutic neutron beam for accelerator-based BNCT. A new beam port assembly was also designed to shape the fast neutrons into epithermal ones and to reduce unnecessary radiation including gammas. As a result of a simulation, it is found that a tumor at a depth of 60 mm from the head skin could be treated within 5 minutes, if a typical tumor is assumed to be taken about 20 RBE Gy for therapeutic treatment. It is, thus, expected that the neutrons emitted into the solid angle bin of 50–150° from 7Li(p,n) 7Be reaction are very effective in producing epithermal neutron beams for BNCT.

Measurements of angular and energy distributions of prompt neutrons from thermal neutron-induced fission

The experimental setup and methodology used to measure prompt neutron angular and energy distributions from thermal neutron-induced fission are described. The neutrons are detected using two scintillation detectors, while the fission fragments are detected by multi-wire proportional detectors in conjunction with the TOF technique. To separate events corresponding to neutrons and γ-quanta, a double discrimination by the pulse shape and the time-of-flight is applied. Some preliminary results of an experiment performed with the 235U target are presented and briefly discussed. The yield of “scission” neutrons has been estimated in the framework of a simple evaporation model and was found not to exceed 5% of the total neutron yield. Including an assumed of anisotropy of the fission neutron angular distribution in the center-of-mass system of fission fragments into the model calculation leads to an increase in the “scission” neutron yields inferred from the data.

Neutron production by a 13C thick target irradiated by 20–90 MeV protons

Neutron production using an enriched 13C carbon converter has been measured during the design study of the italian RIB facility SPES. Energy and angular distributions of neutrons emitted by bombarding a 13C target of stopping length with protons in the range of 20 to 90MeV have been measured by time-of-flight and activation and compared with the prediction of a Monte Carlo code developed at Snezhinsk. At the proton energy of 100MeV, firstly envisaged for SPES, the gain with respect to a natural C target is less than a factor of two, while yields still compare well with those for 40MeV deuterons on natural carbon adopted by SPIRAL-II. At energies near 30MeV the 13C thick target is definitely more prolific than the target of natural carbon, but both yields with protons are clearly lower than the one with deuterons. At the energy of 20MeV envisaged for a first stage of SPES it might be more efficient to irradiate the uranium target with protons rather than using the two-stage method with converter.

Proton-induced fission on actinide nuclei at energies 27 and 63 MeV

Measurements of proton-induced fission on 232Th, 238U, 237Np, 239Pu and 241Am nuclei at proton energies of 26.5 MeV and 62.9 MeV were performed at the Louvain-la-Neuve cyclotron facility. Neutrons were detected with the multidetector DEMON. Mass and energy distributions of fission fragments, neutron angular distribution and multiplicities and total fission cross sections were determined. The relative contributions of pre-equilibrium neutron emission and compound nucleus evaporation, as well as of post-scission evaporation from the fission fragments were analyzed with moving multisource fits and were also compared with the predictions of the nuclear-reaction model code TALYS. Detailed comparisons with the data indicate that there is still room for improvement in this code.

SU‐GG‐T‐333: Neutron Yield and Angular Distribution From the 250 MeV Proton Interactions in Water: A Geant4 Monte Carlo Study

Objective: To simulate using GEANT4 the interaction of 250 MeV protons in water to determine the neutron yield and angular distribution using various hadronic interaction processes. Method and Materials: A cylindrical water phantom (length=40 cm, diameter= 40 cm, density = 1 g/cm3, mean excitation energy = 75 eV) is surrounded by a spherical shell, that consists of 18 detectors, each of which covers 10 degree in theta. In Geant4, electromagnetic energy loss for hadrons and leptons are categorized as either “standard” or “low energy” processes. The low energy process is used to extend down particle interactions with energies below the standard process (for example: Protons 1 keV, electrons and photons 250 eV). The range cut for both low energy and standard process is lowered from the default value of 10−3 m to 1.5×10−5 m to improve the accuracy of simulation. In the case of HI, elastic processes for protons are calculated by G4LEpp and GHEISHA‐style G4LElastic; whereas the inelastic processes for proton and neutron are calculated by G4PreCompoundModel below 170 MeV, and G4CascadeInterface above 150 MeV; and for neutron also by G4NeutronHPorLEInelasticModel. The neutron elastic process is calculated using G4NeutronHPorLEModel for energy greater than 4 eV. When the neutron energy is lower than 4 eV, G4NeutronHPThermalScatteringData is used. The simulations are carried out using 1 million incident pencil beam proton. Results: Neutron fluences per incident proton as a function of theta are determined. Based on hadronic process being used in the calculation, the total number of neutrons per incident proton that reach all eighteen theta detectors ranges from 0.124 to 0.243. Conclusion: Our results show that the hadronic processes indeed play a critical role for GEANT4 simulations. We recommend that careful selection of physics processes for particle interaction is essential in order to interpret realistically simulated results.

SU‐GG‐T‐323: Angular Distribution of Neutron Fluence and Its Effect On Shielding for a Passively‐Scattered Proton Therapy Unit

Purpose: Neutron production is of principal concern in shielding of proton therapy units. Conventionally, neutron calculations are based on the analytical methods, which do not take the actual beam shaping components and the nozzle shielding into account. The goal of this study was to calculate neutron spectral fluence produced at a passive scattering proton treatment unit with detailed Monte Carlo modeling and to estimate the corresponding neutron dose equivalent and shielding requirements at various angles relative to the beam direction. Method and Materials: We modeled all key components in the nozzle including a range modulator, scatter, range shifter and aperture for a passive scattering proton therapy unit with MCNPX. We simulated the spectral neutron fluences at different angles and distances around the nozzle. The neutron spectral fluences were then converted to neutron dose equivalent using conversion factors from International Committee of Radiological Protection (ICRP) Publication 74. Based on the neutron dose equivalent and spectral fluence, the shielding requirements for walls facing the beam and parallel to the beam were assessed. Results: The spectral neutron fluence varied significantly with angle relative to the beam. High energy neutrons were forward peaked and decreased with increasing angle. Low energy neutrons did not change much with angle. Neutron dose equivalent in treatment room did not change significantly with angle, but more shielding was required for the wall facing the beam than other walls because the high energy neutrons were greatest at zero degrees. Conclusion: The neutron spectral fluence from proton therapy units changes significantly with angle relative to the beam direction. Compared to the analytical methods, Monte Carlo simulated results provided a more accurate estimation of the neutron angular distribution for shielding design.

Angular distribution of neutrons from high-intensity laser–target interactions

The angular distribution of neutrons formed in nuclear fusion reactions of a high-energy deuteron beam with a deuterated polyethylene (CD2) was investigated with a Monte Carlo ion beam–target deposition model. The initial conditions were obtained from a two-dimensional particle-in-cell laser–target deposition model. The neutron yield and its angular distribution were studied as a function of peak laser intensity, laser pulse duration and primary target thickness. The proposed scheme for neutron production delivers a typical neutron yield of 10−5–10−3 neutrons/ion and 105–107 neutrons J−1 laser energy.

Neutron production at the heavy ion research facility in Lanzhou

In this work, the neutron radiation field at Heavy Ion Research Facility in Lanzhou (HIRFL) was investigated. Total neutron yields, spectra and angular distributions in the bombardment of various thick targets by 12C and 18O ions with energies up to 75 MeV/u were obtained using the activation method. The neutron dose equivalent rates of 60 MeV/u 18O on various thick targets at different angles were measured with a modified A-B remmeter. Our results are compared with those of other reports.

Neutron-emission mechanisms in the 115In(α, xn) reactions

Spectra and angular distributions of neutrons from the 115In(α, xn) reactions were measured at α-particle energies of 16, 18, 27, and 45 MeV. The measurements were performed with time-of-flight neutron spectrometers at pulsed accelerators of charged particles. The data obtained in this way were analyzed within the models of equilibrium, preequilibrium, and direct reaction mechanisms. The exact formalism of Hauser-Feshbach statistical theory was used in the calculations, level densities in residual nuclei excited in the reactions under consideration being found from the neutron evaporation spectra in the (p, n) reactions on tin isotopes. Contributions from equilibrium, preequilibrium, and direct neutron emission were studied in a wide range of α-particle energies.

Evaluation of D(d,n) 3 He reaction neutron source models for BNCT irradiation system design

A mathematical method was developed to calculate the yield, energy spectrum and angular distribution of neutrons from D(d,n) 3 He (D-D) reaction in a thick deuterium-titanium target for incident deuterons in energies lower than 1.0MeV. The data of energy spectrum and angular distribution were applied to set up the neutron source model for the beam-shaping-assembly (BSA) design of Boron-Neutron-Capture-Therapy (BNCT) using MCNP-4C code. Three cases of D-D neutron source corresponding to incident deuteron energy of 1000, 400 and 150 keV were investigated. The neutron beam characteristics were compared with the model of a 2.45 MeV mono-energetic and isotropic neutron source using an example BSA designed for BNCT irradiation. The results show significant differences in the neutron beam characteristics, particularly the fast neutron component and fast neutron dose in air, between the non-isotropic neutron source model and the 2.5 MeV mono-energetic and isotropic neutron source model.

Comparison and Extension of Dancoff Factors for Pebble-Bed Reactors

A method is developed for the computation of infinite-medium Dancoff factors for spherical kernels with a stochastic packing as used in high-temperature reactors. The method is used to compute Dancoff factors that are then compared to those obtained by four preexisting methods from the literature. The older methods assume either infinitesimally small kernels with a random distribution, or finite kernels with some assumptions. In the new method, the infinite-medium Dancoff factor is calculated rigorously by numerically integrating the Dancoff factor of two adjacent finite-size kernels over their surfaces and relative positions. It turns out that for practical pebble-bed fuel designs, as currently envisioned, all four methods give results accurate (i.e., compatible with one another) within 2%, but that larger deviations are observed for extreme cases (either at high or low dilution).A Monte Carlo program, named INTRAPEB, was written to calculate the average value and the space dependency of the Dancoff factor of one single fuel pebble, as well as the angular distribution of neutrons escaping the pebble. For the Dancoff factor, the analytical results from literature agree very well with the new computational approach. However, for a cubic packing of particles, as is usually modeled in MCNP calculations, a larger Dancoff factor is found for the pebble. The angular distribution of neutrons escaping from the moderator zone of a pebble is much more forwardly peaked than the cosine angular distribution assumed in the derivation of the analytical methods. If the previously developed analytical methods need improvement, this could be achieved by using a more forwardly peaked neutron distribution.A second program, named PEBDAN, was written to calculate the average value and the space dependency of the interpebble Dancoff factor (the probability that a neutron escaping from the fuel zone of a pebble crosses a fuel particle in another pebble) and the pebble-pebble Dancoff factor (the probability that a neutron escaping from the fuel zone of a pebble crosses the fuel zone of another pebble). In this program, the coordinates of the pebbles in a randomly packed bed are determined, after which the Dancoff factors are calculated using a Monte Carlo ray-tracing method. The packing distribution obtained in this code reproduces the principal features of experimental data; however, the predicted radial porosity profile of the packed bed displays less-pronounced oscillations (i.e., lower peaks) and a slightly larger average porosity value. Because of the larger first-flight escape probability of neutrons, the Dancoff factors drop several tens of percent along the inner and outer reflector of the core.

Nuclear break-up of 11Be

The break-up of 11Be was studied at 41 AMeV using a secondary beam of 11Be from the GANIL facility on a 48Ti target by measuring correlations between the 10Be core, the emitted neutrons and gamma rays. The nuclear break-up leading to the emission of a neutron at large angle in the laboratory frame is identified with the “towing mode” through its characteristic n-fragment correlation. The experimental spectra are compared with a model where the time dependent Schrödinger equation (TDSE) is solved for the neutron initially in the 11Be. A good agreement is found between experiment and theory for the shapes of neutron experimental energies and angular distributions. The spectroscopic factor of the 2 s orbital is tentatively extracted to be 0.46 ± 0.15 . The neutron emission from the 1 p and 1 d orbitals is also studied.

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

This article is the first in a series devoted to the development of efficient and accurate computational tools for the design of beam assemblies for boron neutron capture therapy within the framework of discrete ordinates spectral nodal methods of neutron transport theory. We begin our study with a multi-layer representation of an assembly, and we derive a discrete ordinates matrix operator that replaces without spatial truncation error the entire multi-layer domain in neutron transmission computations. With the matrix operator derived here, we compute without further ado the angular distribution of neutrons leaving the multi-layer assembly, avoiding thus the use of general-purpose discrete ordinates codes founded in the discretization and numerical solution of the neutron transport equation over a number of spatial cells and angular directions throughout the domain. We perform numerical experiments with a four-layer model assembly, and we conclude this article with a discussion and directions for further developments.

Angular distribution of fusion products and x rays emitted by a small dense plasma focus machine

Time integrated measurements of the angular distributions of fusion products and x rays in a small dense plasma focus machine are made inside the discharge chamber, using passive detectors. The machine is operated at 37kV with a stored energy of 4.8kJ and a deuterium filling pressure of 2.75torr. Distributions of protons and neutrons are measured with CR-39 Lantrack® nuclear track detectors, on 1.8×0.9cm2 chips, 500μm thick. A set of detectors was placed on a semicircular Teflon® holder, 13cm away from the plasma column, and covered with 15μm Al filters, thus eliminating tritium and helium-3 ions, but not protons and neutrons. A second set was placed on the opposite side of the holder, eliminating protons. The angular distribution of x rays is also studied within the chamber with TLD-200 dosimeters. While the neutron angular distributions can be fitted by Gaussian curves mounted on constant pedestals and the proton distributions are strongly peaked, falling rapidly after ±40°, the x-ray distributions show two maxima around the axis, presumably as a result of the collision of a collimated electron beam against the inner electrode, along the axis.

A nodal SN transport method for three-dimensional triangular- z geometry

A new nodal S N transport method has been developed to perform accurate transport calculation in three-dimensional triangular- z geometry, where arbitrary triangles are transformed into regular triangles via a coordinate transformation. The transverse integration procedure is applied to treat the neutron transport equation in the regular triangle. The neutron angular distributions of intra-node fluxes are represented using the S N quadrature set, and the spatial distributions of neutron fluxes and sources are approximated by a quadratic polynomial. The nodal-equivalent finite difference algorithm for 3D triangular geometry is applied to establish a stable and efficient iterative scheme. The present method was tested on four 3D Takeda benchmark problems published by the nuclear data agency (NEACRP), in which the first three problems are in XYZ geometry and the last one is in hexagonal- z geometry. The results of the present method agree well with those of the reference Monte-Carlo calculation method, the difference in k eff being less than 0.1%. This shows that multi-group reactor core/criticality problems can be accurately and effectively solved using the present method.

Measurements of neutron yield from PF-1000 device by activation method

Silver and indium-detectors have been used since the very beginning of Plasma Focus experiment to measure the total neutron yield and neutron angular distribution. This paper reports on application of other materials to measure neutron yield in Plasma Focus experiments. The obtained results are quite encouraging and the determined neutron emissions are comparable with those obtained by means of the silver activation counters. In order to measure a neutron yield from PF-1000 device an activation method has been applied. Several activation materials have been used including Hf, In, Cd, Ni, Ti and Au. The activity of produced nuclides has been measured after activation allowing to identify those nuclides and γ-lines. Next the respective nuclear reactions have been identified. The neutron yield from PF-1000 has been estimated based on measured activity assuming point, isotropic and mono-energetic (2.45 MeV) source. Each of the foil has been activated during “shots”. These obtained results have been compared with another neutron diagnostics.

Angular distribution of neutrons from deuterated cluster explosions driven by femtosecond laser pulses

We have studied experimentally the angular distributions of fusion neutrons from plasmas of multi-keV ion temperature, created by 40 fs, multi-TW laser pulses in dense plumes of D2 and CD4 clusters. A slight anisotropy in the neutron emission is observed. We attribute this anisotropy to the fact that the differential cross section for DD fusion is anisotropic even at low collision energies, and this, coupled with the geometry of the gas jet target, leads to beam-target neutrons that are slightly directed. The qualitative features of this anisotropy are confirmed by Monte Carlo simulations.