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Instrumental neutron activation analysis (INAA) relies on constant neutron flux densities throughout the activated samples. Although this concept is true for most typical samples, occasionally, the presence of highly neutron absorbing nuclides in the sample may cause a neutron flux density suppression which would ultimately lead to distorted results in the INAA. Here, we have investigated artificial samples with a high manganese (Mn) content. By adding aqueous gold solution, we introduced a liquid in-situ neutron flux monitor into the sample. An Mn content ≤ 50% shows little effect to the internal neutron flux density, however, the flux can be suppressed by ca. 20% when the Mn content reaches 63.2%.

This study presents a dual method for neutron flux self-monitoring and quality control in the k 0 -based neutron activation analysis ( k 0 -NAA) using the Synthetic Multi-Element Standards (SMELS). SMELS comprise three distinct types, designated as Type I for short-lived, Type II for medium-lived, and Type III for long-lived radionuclides, all of which are employed for validating the k 0 -NAA. Each SMELS type incorporates assigned amounts of gold (Au) and can serve as a neutron flux monitor. The Au concentration was automatically integrated into k0-DALAT—a home made software to calculate the gold specific activity ( A sp,Au ), thereby derivating of mass fractions of elements of interest in the k 0 -NAA. Additionally, the assigned concentrations of other elements within SMELS were used for k 0 -NAA quality control purpose. Replicate analyses of NIST-SRM-1633b (Coal Fly Ash) and NIST-SRM-679 (Brick Clay) by k 0 -NAA showed determination up to 29 and 25 elements, respectively with deviations within 7%, of which u -scores were almost less than ∣2.58∣, except only for Na and Fe of all 2.99. It revealed that the dual method using SMELS for both neutron flux self-monitoring and quality control considered convenient and consistent in comparison with the results obtained by the conventional independent comparator method biased within 1%.

. A new high flux experimental area has recently become operational at the n_TOF facility at CERN. This new measuring station, n_TOF-EAR2, is placed at the end of a vertical beam line at a distance of approximately 20m from the spallation target. The characterization of the neutron beam, in terms of flux, spatial profile and resolution function, is of crucial importance for the feasibility study and data analysis of all measurements to be performed in the new area. In this paper, the measurement of the neutron flux, performed with different solid-state and gaseous detection systems, and using three neutron-converting reactions considered standard in different energy regions is reported. The results of the various measurements have been combined, yielding an evaluated neutron energy distribution in a wide energy range, from 2meV to 100MeV, with an accuracy ranging from 2%, at low energy, to 6% in the high-energy region. In addition, an absolute normalization of the n_TOF-EAR2 neutron flux has been obtained by means of an activation measurement performed with 197 Au foils in the beam.

To treat superficial tumors using accelerator-based boron neutron capture therapy (ABBNCT), a technique was investigated, based on which, a single-neutron modulator was placed inside a collimator and was irradiated with thermal neutrons. In large tumors, the dose was reduced at their edges. The objective was to generate a uniform and therapeutic intensity dose distribution. In this study, we developed a method for optimizing the shape of the intensity modulator and irradiation time ratio to generate a uniform dose distribution to treat superficial tumors of various shapes. A computational tool was developed, which performed Monte Carlo simulations using 424 different source combinations. We determined the shape of the intensity modulator with the highest minimum tumor dose. The homogeneity index (HI), which evaluates uniformity, was also derived. To evaluate the efficacy of this method, the dose distribution of a tumor with a diameter of 100 mm and thickness of 10 mm was evaluated. Furthermore, irradiation experiments were conducted using an ABBNCT system. The thermal neutron flux distribution outcomes that have considerable impacts on the tumor’s dose confirmed a good agreement between experiments and calculations. Moreover, the minimum tumor dose and HI improved by 20 and 36%, respectively, compared with the irradiation case wherein a single-neutron modulator was used. The proposed method improves the minimum tumor volume and uniformity. The results demonstrate the method’s efficacy in ABBNCT for the treatment of superficial tumors.

The distribution of the thermal neutron flux has a significant impact on the treatment efficacy. We developed an irradiation method of overlapping irradiation fields using intensity modulators for the treatment of superficial tumors with the aim of expanding the indications for accelerator-based boron neutron capture therapy (BNCT). The shape of the intensity modulator was determined and Monte Carlo simulations were carried out to determine the uniformity of the resulting thermal neutron flux distribution. The intensity modulators were then fabricated and irradiation tests were conducted, which resulted in the formation of a uniform thermal neutron flux distribution. Finally, an evaluation of the tumor dose distribution showed that when two irradiation fields overlapped, the minimum tumor dose was 27.4 Gy-eq, which was higher than the tumor control dose of 20 Gy-eq. Furthermore, it was found that the uniformity of the treatment was improved 47% as compared to the treatment that uses a single irradiation field. This clearly demonstrates the effectiveness of this technique and the possibility of expanding the indications to superficially located tumors.

This paper introduced a compact high flux polarized neutron beam generator scheme, which used air as the working medium and had low energy consumption. The neutron beam generator adopted a linear three compartment configuration, sequentially using nitrogen nucleus tandem near range accelerated polarization target spallation nuclear reaction technology, neutron multiplication technology, neutron beam polarization and near range acceleration technology, neutron focusing and shooting control technology. Through design and equivalent verification, it has been proven that the total length of the device does not exceed 5 m, the effective range can reach several hundred kilometers, the neutron flux at the muzzle is not less than 1025 n·cm−2·s−1, which attenuates to 1010 n·cm−2·s−1 at a distance of several 100 km, and this flux can effectively strike the target. It can be used as a defensive directed energy weapon with high energy density and has broad application prospects.

The article presents an experimental confirmation of the operability of neutron concentrators in devices that form and use directed high-intensity thermal neutron beams with elliptical channels made as blocks of profiled graphite and aluminum plates. The effect of neutron reflection from the surface of materials is the basis of a device capable of selecting neutrons by their directions in space. The study experimentally confirmed the efficiency of a moderating-focusing structure (MFS) based on a pack of elliptical neutron mirrors, which makes it possible to form oriented thermal neutron beams from the outgoing neutron flux. To record the effects of selective thermal neutron separation, silicon single-crystal wafers were used, due to which it was possible to obtain portraits of integral neutron fluxes in the reactor. The experiments were carried out in a horizontal experimental channel (HEC-4) at the IRT-T reactor of the National Research Tomsk Polytechnic University. The integral neutron flux was (2.3–3.02)·10 17 cm –2 . The neutron flux was detected by the change in the specific electrical resistivity of the single-crystal silicon wafers. The effect of concentration of thermal neutrons was recorded both on the block of graphite neutron mirrors and on the block of aluminum thin-walled elliptical mirrors. In the near future, on this basis, it will be possible to solve such problems as extending the reactor life by reducing the hydrogen uptake in the inner walls. In addition, the experiments have proved the possibility of creating anisotropic structures that lie outside the formalism of Liouville’s theorem, in which the surfaces of thermal neutron sinks are formed with subsequent concentration in the areas separated by aluminum or graphite plates.

Design of Irradiation Facilities at Grid E-1 of Plate Type Research Reactor Bandung. Plate Type Research Reactor Bandung (PTRRB) core design is one of the result of PTRRB research programs. In the previous study the irradiation facilities at grid E-1 has not been designed and also distribution of thermal, epithermal and fast neutron flux at grid E-1 has not been studied. Since that data is very important especially in radioisotope production and neutron beam tube analysis, therefore in this study irradiation facilities at grid E-1will be designed. Previous PTRRB core design is a base for designing irradiation facilities at grid E-1. Considering geometrical of grid E-1 and aluminum tube dimension there are three possibilities aluminum tube configuration. The configurations are configuration 1, 2 and 3. Each configuration was modelled as arrangement of four aluminum tubes and each tube filled by four aluminum irradiation capsules. That configuration was starting point to made MCNP PTRRB reactor core model so there are three MCNP PTRRB reactor core model. MCNP PTRRB reactor core model is needed because MCNP software are computer program for calculating excess reactivity and neutron flux distribution at grid E-1. Result excess reactivity calculation of three configuration indicate that after installing irradiation tube excess reactivity is lower than of limit excess reactivity value 10.9 % of neutronic safety criteria of PTRRB design. Based on neutronic safety criteria, the three configuration is accepted for irradiation facilities PTRRB. Neutron flux calculation result of three configuration reveals that the highest neutron flux is located at capsule no II and III. Profile of thermal neutron flux, epithermal neutron flux and fast neutron flux of three configurations are similar. Neutron flux of thermal, epithermal and fast neutron of three configuration are slightly different. The calculation result reveal that highest thermal neutron flux at grid E-1 is 2.70 × 1013(n/cm2.sec) at configuration 2. Based on neutronic safety criteria and thermal neutron flux, configuration 2 is appropriate for irradiation facilities of PTRRB.

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution.