Neutron Filter Research Articles

Production of Optimal Epithermal Neutron Beams for BNCT

Boron Neutron Capture Therapy (BNCT) is a potential treatment for malignant brain tumors. Depending on the position and size of the tumor, the necessary energy, intensity and resolution of the incident neutron beams are required. To minimize collateral effects on healthy tissues, specific neutron filters were estimated in order to produce optimal epithermal neutron beams. The optimal components and amounts of elements for such filters were calculated. Therapeutic neutron beams, with high spectral purity, were produced.

Dual Spectrum Neutron Radiography: Identification of Phase Transitions between Frozen and Liquid Water

In this Letter, a new approach to distinguish liquid water and ice based on dual spectrum neutron radiography is presented. The distinction is based on arising differences between the cross section of water and ice in the cold energy range. As a significant portion of the energy spectrum of the ICON beam line at Paul Scherrer Institut is in the thermal energy range, no differences can be observed with the entire beam. Introducing a polycrystalline neutron filter (beryllium) inside the beam, neutrons above its cutoff energy are filtered out and the cold energy region is emphasized. Finally, a contrast of about 1.6% is obtained with our imaging setup between liquid water and ice. Based on this measurement concept, the temporal evolution of the aggregate state of water can be investigated without any prior knowledge of its thickness. Using this technique, we could unambiguously prove the production of supercooled water inside fuel cells with a direct measurement method.

Progress of Filtered Neutron Beams Development and Applications at the Horizontal Channels No.2 and No.4 of Dalat Nuclear Research Reactor

The neutron filter technique has been applied to create mono-energetic neutron beams with high intensity, at the horizontal channels No.2 and No.4 of the Dalat nuclear research reactor. The mono-energetic neutron beams that have been developed for researches and applications are thermal (0.025eV), 24keV, 54keV, 59keV, 133keV and 148keV. The relative intensities of main peak in filtered neutron energy spectra and the collimated neutron fluxes at the sample irradiation positions are 90 ¸ 96% and 2.8×105 ¸ 7.8×106 n/cm2.s, respectively. Monte Carlo simulations and transmission calculations were performed to each neutron energy beam for optimal design of geometrical structure and neutron filter materials. These filtered neutron beams have been applied efficiently for experimental researches on neutron total and capture cross sections measurements, and elemental analysis in various kinds of samples based on the prompt gamma neutron activation analysis method. This paper reviews the progress of filtered neutron beams development and its applications for past many years at the Dalat nuclear research reactor.

Design calculation of a horizontal thermal neutronic beam for neutron radiography at the Syrian MNSR

The computer code MCNP4C and the ENDF/B-V cross-section library were used to design calculation of a horizontal thermal beam for neutron radiography (NR) at Syrian MNSR and to evaluate the safety of the reactor after installation of the NR facility (NRF). Thermal, epithermal and fast neutron energy ranges were selected as 10.0 keV, respectively. To produce a good neutron beam in terms of intensity and quality, bismuth (Bi) and silicon (Si) were used as photon and neutron filters, respectively. The ratio of L/D of the NRF ranges between 90 and 125. The thermal neutron flux at the beam exit plane can be varied from 1.836 × 105 to 3.057 × 105 n/cm2 s. If such thermal neutron beam would be built into the Syrian MNSR, many scientific applications of the NR would be available.

Design of an epithermal neutron beam for BNCT in thermal column of Tehran research reactor

The potential of the thermal column of Tehran Research Reactor (TRR) to provide epithermal neutron beam for Boron Neutron Capture Therapy (BNCT) has been investigated using Monte Carlo simulation. A Beam Shaping Assembly (BSA) has been designed and optimized to meet BNCT neutron beam criteria recommended by International Atomic Energy Agency. The suggested BSA configuration in cylindrical geometry consists of 20cm Al as a moderator, 35cm Pb as a reflector, two 5cm Bi slabs as gamma shield, and two 2mm Cd sheets as thermal neutron filters. The results show that epithermal neutron flux at the exit of the BSA can be 0.65×109n/cm2s. In-phantom dose analysis indicates that the designed neutron beam can be used for treatment of deep-seated brain tumors in acceptable time.

Development of 24 and 59 keV Filtered Neutron Beams for Neutron Capture Experiments at Dalat Research Reactor

External filtered neutron beams have been developed at the horizontal radial channels No. 4 of Dalat research reactor. In the material composition of the neutron filters, the primary material components of Iron, Aluminum, Nickel and Vanadium are used to obtain the mono-energetic neutron beams of 24 and 59 keV, with low level of Gamma and slow neutron background. A computer code and Monte-Carlo simulation technique were applied to optimize the filter configurations and to deduce the neutron energy distributions in the filtered beams. A hydrogen-filled proton recoil detector and the activation method with Gold foils were used to measure the neutron energy spectrum and flux of each beam at sample position. The results of experimental neutron fluxes are 6.1 × 105 and 5.3 × 105 n/cm2/s for 24 and 59 keV beams, respectively.

Distinction of Liquid Water and Ice Based on Dual Spectrum Neutron Imaging

This study investigates a new approach to distinguish liquid water and ice based on dual spectrum neutron imaging. The imaging technique benefits from arising differences between the cross section of water and ice at neutron wavelengths above 3 Å. As a significant portion of the energy spectrum of the ICON beamline at PSI is below 3 Å, no differences can be observed with the entire beam. However, introducing a Beryllium neutron filter inside the beam, the interesting region of the spectrum is emphasized resulting in a significant contrast of approximately 1.8% between the aggregate states of liquid and frozen water.

Optimization of the beam shaping assembly in the D–D neutron generators-based BNCT using the response matrix method

Optimization of the Beam Shaping Assembly (BSA) has been performed using the MCNP4C Monte Carlo code to shape the 2.45MeV neutrons that are produced in the D–D neutron generator. Optimal design of the BSA has been chosen by considering in-air figures of merit (FOM) which consists of 70cm Fluental as a moderator, 30cm Pb as a reflector, 2mm 6Li as a thermal neutron filter and 2mm Pb as a gamma filter. The neutron beam can be evaluated by in-phantom parameters, from which therapeutic gain can be derived. Direct evaluation of both set of FOMs (in-air and in-phantom) is very time consuming. In this paper a Response Matrix (RM) method has been suggested to reduce the computing time. This method is based on considering the neutron spectrum at the beam exit and calculating contribution of various dose components in phantom to calculate the Response Matrix. Results show good agreement between direct calculation and the RM method.

Neutron characteristics of single-crystal magnesium fluoride

A simple formula is proposed that permits the calculation of both the neutron reflectivity and attenuation of single-crystal magnesium fluoride (MgF2) as a function of wavelength at both room temperature and 77K. A computer program MgF2 written in FORTRAN-77, has been developed to carry out the required calculations for the single-crystal MgF2, when it is used as a neutron monochromator and/or filter.The monochromatic characteristics of single-crystal MgF2 are detailed in terms of the optimum crystal cutting plane, mosaic spread, thickness and reactor moderating temperature for efficient neutron reflectivity within the wavelength band from 0.1nm up to .45nm. Calculation shows that a 7mm thick MgF2 single-crystal cut along (111) plane having 0.50 FWHM on mosaic spread has the optimum parameters when it is used as a monochromator at neutron wavelengths shorter than 0.2nm. However, the integrated neutron intensity of 2nd and 3rd orders from a thermal reactor flux is even higher than that of the 1st order one at neutron wavelengths longer than 0.2nm. While, from a cold reactor flux, the use of the single-crystal MgF2 (110) as a neutron monochromator free from higher order contaminations at λ⩾0.38nm is more appropriate.A feasibility study of using single-crystal MgF2 as a neutron filter is also detailed in terms of the optimum crystal orientation, mosaic spread, thickness and temperature for efficient transmission of thermal-reactor neutrons. Calculation shows that a 100mm thick MgF2 single-crystal cut along (110) plane cooled to liquid nitrogen temperature, with 0.05o FWHM on mosaic spread, is a good thermal neutron filter, with high effect-to-noise ratio.

A new irradiation method with a neutron filter for silicon neutron transmutation doping at the Japan research reactor no. 3 (JRR-3)

We studied a new silicon irradiation holder with a neutron filter designed to make the vertical neutron flux profile uniform. Since an irradiation holder has to be made of a low activation material, we applied aluminum blended with B4C as the holder material. Irradiation methods to achieve uniform flux with a filter are discussed using Monte-Carlo calculation code MVP. Validation of the use of the MVP code for the holder's analyses is also discussed via characteristic experiments.

Study on the burn-up characteristics of a thermal neutron filter containing B4C particles for NTD-Si irradiation

We investigated an alternative irradiation method for silicon doping using a thermal neutron filter to increase the irradiation efficiency in the Japan Research Reactor No. 3 (JRR-3). Aluminum mixed B4C particles were adopted as the filter material. The burn-up characteristics of a thermal neutron filter containing a large number of B4C particles are described. We have developed a method for evaluating neutron transmissivity as Boron burning in the neutron filter and clarified the applicable scope of the method via experiments using thermal neutron filters. We derived a method by which we could calculate the suitable diameter and density of B4C particles.

A dynamically adjustable wavelength-sensitive neutron filter

The prototype of a wavelength-sensitive neutron filter has been realized and tested successfully for the first time. The filter exploits the neutron wavelength dependence of the transparency of 3He gas. The analysis of test results gives high wavelength resolution Δλ/λ≃0.5% over the broad wavelength range 0.2–5.2Å. We propose an idea to expand the filter range in the short wavelength direction by use of a cryogenic environment, and also consider the use of media using other than 3He gas. The proposed filter might be used in advanced reflectometers and SANS2D neutron scattering instruments.

Evaluation of Design Neutron Filters in BNCT

7 neutrons per Second per milliampere of electron current is shown to be obtainable. The useful neutron energy range for BNCT purposes is from 1 eV to 10 keV (sometimes 40 keV is as-summed). The effectiveness of the treatment is expected to increase both the amount of boron in the cells and the neutron flux increase. In this treatment, advantage is taken of the nuclear reaction between boron and thermal neutrons, resulting in cell death from the energy deposited along the paths of resultant alpha particles and lithium ions. Abstract This paper reviews the development of neutron filters using Boron Neutron Capture Therapy (BNCT). At present BNCT research necessity of an epithermal beam is to generate the necessary thermal neutron field at the desired depth. Through the use of an epithermal beam, deeper-seated tumors can be treated more effectively (1). Epithermal neutron beams can be generated by small nuclear reactors and by accelerator based neutron sources. So far, only reactors have been actually used to produce therapeutically useful epithermal neutron beams for BNCT; accelerator- based epithermal neutron sources may constitute the basis for a more deployable technology for BNCT than reactor- based sources in the long term (2). Accelerators offer a number of potential advantages over reactor-based neutron sources for clinical applications (3). The goal of reactor epithermal neutron filter conceptual design effort was to provide a low cost, simple filter design that is easy to fabricate and install while meeting or exceeding the nuclear performance requirements of the BNCT program. Therefore, in this work the use of aluminum compound, teflon compound and fluental and heavy water was considered.

QMENF-G: A computer package for quasi-mono-energetic neutron filters

A computer package QMENF-G was developed in a general form to calculate the optimum amounts and components of quasi-mono-energetic neutron filters and their transmitted spectra in order to meet requirements for different applications in various fields. This code is a complete system to assist the user with input preparation, running calculations and analysis of output results. It is written in standard FORTRAN that runs on any personal computer using windows operating system. As a test case, the components and amounts of elements for producing a neutron filter at ∼3.8 keV was calculated using QMENF-G code. It was found that 54Fe, 80Se or 90Zr, can be used as the main components of such filter. However, its transmission, resolution and purity were found to depend upon the accompanying components and their amounts.

Design of a model for BSA to meet free beam parameters for BNCT based on multiplier system for D–T neutron source

Extensive research has recently been carried out for the development of high-energy D–T neutron generators as neutron sources for BNCT. The energy of these high-energy neutrons must be reduced by designing a Beam Shaping Assembly (BSA) to make them usable for BNCT. However, the neutron flux decreases drastically as neutrons pass through different materials of BSA. Therefore, it is very important to find ways to treat the neutrons economically. In this paper the possibility of using natural uranium as a neutron multiplier is investigated in order to increase the number of neutrons emitted from D–T neutron generator. According to the simulations and performed calculations, a sphere containing natural uranium as neutron multiplier was used to increase the number of neutrons generated by the D–T neutron generator. The energy of fast neutrons that are generated by D–T fusion reaction and amplified by neutron multiplier system is decreased using proper materials as moderators and fast neutron filters in BSA. The gamma rays which are generated as a result of neutron interaction with moderators are removed from neutron spectrum using bismuth as the gamma filter. Also, a thermal neutron absorber omits undesired low-energy neutrons which lead to a high radiation dose for the skin and soft tissues. The results show that passing neutrons through such a BSA causes the establishment of free beam parameters yet the reduction of the output beam intensity is unavoidable. The neutron spectrum related to our BSA has a proper epithermal flux and the fast and thermal neutron fluxes are compatible with the IAEA recommended values.

MgO single-crystal as an efficient thermal neutron filter

Abstract A formula is given which, for neutron energies in the range 10 −4 E Calculation shows that 10 cm thick MgO-single crystal cut along its (1 0 0) plane, with 0.5° FWHM on mosaic spread and cooled at 77 K, is a good thermal neutron filter, with high effect-to-noise ratio.

Characterization of a new external neutron beam facility at the Ohio State University

A thermal neutron beam facility has been designed and implemented at the Ohio State University Research Reactor. A project is underway to construct a large vacuum chamber such that the facility could have neutron depth profiling and neutron radiography capabili- ties as intended. The neutron beam is extracted from the reactor through a neutron collimator emplaced in Beam Port #2. The neutron spectrum entering the neutron colli- mator was unfolded from foil activation analysis results and also simulated with a full reactor core model in the MCNP Monte Carlo code. The neutron collimator uses polycrystalline bismuth as a gamma ray filter and single- crystal sapphire as a fast neutron filter. The beam is defined by multiple 3.0 cm diameter apertures made of borated aluminum. Characterization of the beam was performed using foil activation to find the flux and a low-budget neutron imaging apparatus to see the beam profile. The modulation transfer function was calculated to offer insight into the resolution of the imaging system and the colli- mation of the beam. The neutron collimator delivers the filtered thermal neutron beam with a *4 cm diameter and a thermal equivalent flux of (1.27 ± 0.03) 9 10 7 n/(cm 2 s) at 450 kW power at the end of the collimator.

Detailed Design of 242mAm Breeding in Pressurized Water Reactors

There is growing interest in the use of 242m Am as a nuclear fuel. Because of its very high thermal fission cross section and its large number of neutrons released per fission, it can be used for various unique applications, such as space propulsion, medical applications, and compact energy sources. Since the thermal absorption cross section of 242mAm is very high, the best way to obtain 242mAm is by the capture of fast or epithermal neutrons in 241 Am. However, fast spectrum reactors are not readily available.In this paper, we explore the possibility of producing 242mAm in existing pressurized water reactors (PWRs) with minimal interference in reactor performance. As suggested in previous studies on the subject, the 242mAm breeding targets are shielded with strong thermal absorbers in order to suppress the thermal neutron flux that causes 242mAm destruction. Since 242mAm enrichment within the Am target mainly depends on the neutron energy distribution, which in turn depends on the Am target thickness and on the neutron filter cutoff energy (thermal absorber type), this unique Am target design was developed.In our study, Cd, Sm, and Gd were considered as thermal neutron filters, as suggested by Cesana et al. The most favorable results were obtained by irradiating Am targets covered either with Gd or Cd. In these cases, up to 8.65% enrichment of 242mAm is obtained after 4.5 yr (three successive PWR fuel cycles) of irradiation. It was also found that significant quantities [up to 1.3 kg/GW(electric)⋅yr] of 242mAm can be obtained in PWR reactors without notable interference with reactor performance. However, in order to maintain the original fuel cycle length, the enrichment of the driver (UO2) fuel must be increased by ˜1%, raised from the conventional 4.5 to 5.5%, depending on the thermal neutron filter used. The most important reactivity feedback coefficients for fuel assemblies containing the 242mAm breeding targets were evaluated and found to be close to those of a standard PWR.Another product of neutron capture in the 241Am reaction is 238 Pu. It was found that in a typical 1000 MW(electric) PWR core with one-third of the fuel assemblies containing 241 Am targets, up to 15.1 kg of 238Pu enriched to 80% can be produced per year.

GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities. In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]). The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×1011 ncm2/s. This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent–divergent collimator geometry has been adopted. The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line (http://www.info.cern.ch/asd/geant4/geant4.html [4]). To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al2O3) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al2O3) filters, respectively. To get a good cadmium ratio, GEANT 4 simulations were used to define the design of the moderator in the inlet of the radiation channel. A graphite block of 22 cm thickness seems to be the optimal neutron moderator. The results showed that the combination of 5 cm of bismuth with 5 cm of sapphire permits the filtration of gamma-rays, epithermal neutrons as well as fast neutrons in a considerable way without affecting the neutron thermal flux.

Implementation and testing of the JET gamma-ray cameras neutron filters pneumatic system

Neutron filters/attenuators have been designed and constructed as an upgrade of the JET gamma-ray cameras (GRC's). This diagnostics upgrade should reduce the neutron flux at the gamma-ray detectors in a way that it would make possible gamma-ray imaging measurements in high power deuterium JET pulses, and eventually in deuterium–tritium discharges. The attenuators will be placed within the GRC diagnostics system between the vacuum vessel port and the camera collimator radiation shields in the case of both the horizontal camera and the vertical one and they have to be moved out of the detector line of sight when the neutron/gamma camera diagnostics is used for neutron measurements. The neutron filters will operate in a very harsh electromagnetic environment. It is thus recommended to avoid as much as possible the use of electrical and electronic components for the attenuator steering and control (in particular, for attenuator steering). Therefore a steering and control system based on pneumatic components was designed and constructed. The attenuators together with the steering and control system have been installed on a test stand, which is a 1–1 scale replica of the JET structures where neutron filters are to be installed and successfully tested.