Fast Neutron Flux Research Articles

Neutron spectrum determination of accelerator-driven d(10)+Be neutron source using the multi-foil activation technique

New neutron field of the compact accelerator-driven fast neutron source based on the cyclotron and beryllium target station was studied and developed at the NPI CAS. Neutrons were generated using the deuteron bombardment on 8 mm thick beryllium layer by the 10.3 MeV deuteron beam. Broad neutron energy spectra of the d(10)+Be source reaction at two source-to-sample distances (14 mm and 154 mm) were determined for the first time at the NPI. Sets of eight activation materials comprising Au, Co, Ti, In, Al, Fe, Ni, and Nb were irradiated in the d(10)+Be neutron field, and activated dosimetry foils were analysed using the nuclear gamma-spectroscopy technique (HPGe detector) and reaction rates were measured. Neutron energy spectra of the d + Be source reaction for 10.3 MeV deuterons were reconstructed employing the modified version of the SAND-II unfolding code. For 9.7 μA deuteron beam, the fast neutron flux of developed d(10)+Be neutron field reached the value of 1.4 × 1010 cm−2s−1 at a distance of 14 mm from the source target and 4.1 × 108 cm−2s−1 at a distance of 154 mm. The novel d(10)+Be neutron field with energy range up to 15 MeV is convenient for neutron cross-section data validation, experimental simulation of the fast neutron spectrum of experimental nuclear reactors, applications of neutron activation analysis, and radiation hardness tests of materials important for nuclear energetics.

The dependency of fast neutron flux and spectra on the collimation-field size and geometry for fast neutron therapy purposes

The aim of this work is to optimize and investigate the relationship between the collimation-filed size and their geometries and the fast-neutron intensity and their energy-spectrum. Advantageously for these purposes, a prototype of collimator was designed and constructed to simulate different geometrical and materials compositions. Beams of fast neutrons generated by the reaction 9Be(d, n) with 13.6 MeV deuteron energy were investigated. The average energy of the emitted neutrons is about 6–7 MeV. The maximum intensity was at neutron energy of 4–5 MeV. In the process of the study, experimental works were carried out with Al, Fe, Cu and Cd foils which positioned on the collimator aperture as neutron detectors. The neutron activation method and Gamma-spectrometer were deployed to determine the neutron intensities within the energy-range from thermal energies to 14 MeV. Experiments were conducted for ten different combinations of materials and geometries which can be adjusted by the removable-polyethylene collimator parts and the metal parts. It was concluded that the thermal and fast neutron flux were depended on the collimation-field size and materials, where the fast neutron flux output was about 100% more for bigger collimator 9 × 9 cm2 comparing with the smaller ones 5 × 5 cm2 . And more than 60% in case of adding Lead layers into the collimator. These results could have influential implications on the fast neutron intensities and improvement of collimation systems for fast neutron therapy. Also, it could enhance the irradiation fluxes in the irradiation channels in nuclear reactors for medical isotopes productions and material testing and other applications.

Solution for the Multigroup Neutron Space Kinetics Equations by Source Iterative Method

In this work, we used a modified Picard’s method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a first order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform by Stehfest method. We present numerical simulations and comparisons with available results in literature.

Integral experiments of technetium-99 using fast-neutron source reactor ‘YAYOI’

ABSTRACT The present study performed integral experiments of 99Tc by an activation method using a fast-neutron source reactor ‘YAYOI’ of the University of Tokyo to validate evaluated nuclear data libraries: JENDL-4.0, ENDF/B-VII and VIII. Technetium-99 samples were irradiated with reactor neutrons using a pneumatic system, and repeated 74 times to accumulate statistics. Reaction rates of 99Tc were obtained by measuring decay gamma rays emitted from 100Tc. The fast-neutron flux spectrum at the YAYOI was checked with flux monitors and MCNP calculations. The experimental reaction rate of 99Tc was compared with the reaction rates given using both the fast-neutron flux spectrum and evaluated nuclear data libraries. The present integral experiments supported the evaluated nuclear data library JENDL-4.0.

Fullerite Nano-materials as a Moderator in Neutron Irradiators with Radium Sources: Feasibility and Advantages

Abstract: In this study, the feasibility of using fullerite nano-materials as a moderator in 226Ra-Be neutron irradiators has been theoretically investigated, for the first time. Thermal, intermediate and rapid neutron flux in irradiation channels was calculated using the MCNP5 code when a fullerite nano-material was used as a moderator. The simulation results were then compared with other simulation results performed when paraffin was used as a moderator. The comparison showed that using fullerite instead of paraffin as a moderator results in an increase in the total number of irradiation neutrons by more than twice in average (240 %) for each direction inside the irradiator. This increase is distributed as follows: 27.84 %, 87.84 % and 124.32 % thermal, intermediate and rapid neutrons, respectively. The previous distribution indicates a significant increase in the intermediate and fast neutron flux. This is considered as an additional advantage of using the 226Ra-Be neutron irradiator with a fullerite moderator. The irradiator can then be used not only to irradiate the materials whose irradiation requires thermal neutrons, but also those that require medium- to high-energy neutrons. Keywords: Neutronic irradiator, Ra-Be radiation source, Cadmium, Neutron flux, MCNP5-beta code. PACS: Neutrons diffusion and moderation, 28.20.Gd, Moderators (nuclear reactors), 28.41.Pa.

Neutron guide optimization for the Moroccan PGAA system

The Moroccan TRIGA Mark II research reactor is to be equipped with a PGAA (Prompt Gamma Activation Analysis) facility to assist the country's progression in socioeconomic areas such as the environment and geochemistry, agriculture, health, industry, cultural heritage, and human sciences. The requirements of the PGAA facility include a very high thermal to fast neutron flux ratio and a focused, thin beam. Supermirror neutron guides are generally used to meet such requirements due to their ability to reflect neutrons with a specific energy and scattering angle. This study was undertaken to determine a suitable neutron guide for achieving optimal performance for the PGAA facility, where such a determination needs to be made according to both the available space and the minimum performance needed for PGAA analysis. To better parameterize the neutron guide, a set of simulations were performed with the aim of establishing the best guide specifications. Three types of neutron guides were studied, namely straight, curved, and bender. McStas code was used in this study, and the simulations showed that the guides must be 6m long and that the coating value m must be equal to 5 for the bender and straight guides and 6 for the curved guide. Among the three guide types, the thermal/epithermal (Фth/Фep) and thermal/fast (Фth/Фfast) ratios were found to be much better when using the curved guide.

The low enriched uranium miniature neutron source reactor (LEU-MNSR) neutron spectrum characterization for k0-INAA

Ghana’s highly enriched uranium (90.2%) research reactor has been converted to a low enriched uranium (13.0%) type. Its neutron parameters have been characterized for the application of k 0 -instrumental neutron activation analysis. The magnitude of the LEU epithermal neutron shaping factor ( α ) at the irradiation site increased by about 25% as compared to the HEU, while the thermal neutrons decreased by about 5%. This indicates that the epithermal neutron spectrum is much dependent on the 1/ E 1 + α . Similarly, the measured fast neutron flux increased by about 22%. Generally, the results show that the LEU core’s neutron spectrum is relatively harder, i.e., there is a marginal increase in high-energy neutrons. The reactor spectrum parameters have been used to validate the k 0 -IAEA software. Two certified reference materials (Carbonate rock GBW 07108 (GBW2) and NIST Standard Reference Material 1646a, Estuarine Sediment (EST)) and the synthetic multi-element standard (SMELS I, II & III) were analyzed based on a routine irradiation scheme. The process quality assurance was evaluated based on a u-score test. Most (80 %) of the elements determined were within | u-score | < 1.64, while a few were observed within 1.964 > | u-score | > 1.64. The results show a suitable characterization process and the general applicability of the k 0 -IAEA software. • Neutron spectrum at irradiation sites of MNSR with LEU core characterized. • Neutron spectrum for MNSR with LEU core harder as compared to the HEU core. • K 0 -IAEA software calibrated using measured neutron spectrum parameters. • Good agreement between certified and measured concentrations using the k 0 -IAEA software indicate validity of measured neutron spectrum parameters.

Distributed Fiber Sensors With High Spatial Resolution in Extreme Radiation Environments in Nuclear Reactor Cores

This paper is a comprehensive experimental report on the neutron radiation effects of distributed optical fiber sensors with enhanced Rayleigh scattering profiles in an in-pile environment. Femtosecond laser direct writing was used to inscribe Type-II modifications in standard telecom fibers and radiation-hardened fibers with fluorine-doped cores. Rayleigh backscattering signals were enhanced for continuous 1.5 m. In-pile lead-out sensors tests were carried out at the MIT Research Reactor for two months, which was operated at a nominal power of 5.7 MW with fast neutron (>0.1 MeV) flux of 1.29 × 10 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">14</sup> n/cm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> /s and an in-core temperature of up to 560 °C. Using the Optical Frequency Domain Reflectometry technique, the backscattering profiles of fiber sensors were interrogated with a 3-cm spatial resolution to monitor the temperature profile of the reactor. Results show that laser inscribed Type-II modifications in the form of nanogratings are highly stable against extreme temperature and ionizing radiation. Both standard telecom fibers and radiation-hardened fibers with laser-enhanced Rayleigh profiles can continuously perform distributed temperature measurements over the entire duration of the in-pile testing. Temperature coefficients of sensors and spectral shift quality were studied as functions of total radiation fluence. To the best of our knowledge, we present for the first time, the temperature profile of an operating nuclear reactor core with 3-cm spatial resolution, enabled by distributed fiber sensors with laser-enhanced Rayleigh scattering profiles. The high spatial resolution measurements can provide valuable data for the design and validation of digital twin and virtual reality of nuclear energy systems.

Microstructure and hardening induced by neutron irradiation in single crystal, ITER specification and cold rolled tungsten

In this work, we have performed neutron irradiation and sub-sequent microstructural and micro-hardness measurements on a series of tungsten grades. The neutron irradiation was performed in BR2 reactor up to 0.2 dpa at 600, 800 and 1200 °C. The selected irradiation temperatures correspond to a large range of the operation of tungsten in ITER as armour for the divertor. The applied fast neutron flux is representative of ITER irradiation conditions in terms of damage rate. The primary purpose of the study was to establish the relation between the initial microstructure, the resulting irradiation-induced microstructure and corresponding irradiation hardening. The microstructure was studied using transmission electron microscopy (TEM) and hardening was assessed by micro-hardness measurements. The obtained TEM results were used to predict the hardening employing the dispersed barrier model and strength coefficients established for the single crystal from the open literature. Based on the performed analysis, it appeared that the main contribution to the hardening at high irradiation temperature originates from the voids, while the dislocation loops provide a comparable contribution (30–50%) only under irradiation at 600 °C. Excellent agreement between the model prediction and experimentally measured hardness increase was obtained for the single crystal and ITER specification grade without any additional adjustment, while the hardening induced in the cold rolled plate was essentially overestimated by the model. This result suggests that the cold rolled tungsten plate exhibits an alternative mechanisms of the plastic deformation besides the conventional dislocation glide and multiplication, and this alternative mechanism is impacted by the irradiation less severely compared to the prediction coming from the dispersed barrier model. • Neutron irradiation of tungsten at 1200 °C still leads to pronounced hardening. • Voids are the main source of hardening after irradiation in the temperature range 600–1200 °C. • Irradiation causes significant reduction of initial dislocation density in cold rolled tungsten.

Design and optimization of thermal neutron device based on deuterium-deuterium neutron generator

• A thermal neutron device has been designed and optimized using the MCNP5 code to increase thermal neutron flux. • The system performance is evaluated using two parameters P and K. • The thermal neutron flux is 2.53 × 10 3 n/cm 2 s, accounting for 39.5 % of the total neutron flux on the output surface. • The system can be used to neutron activation analysis, test and calibrate neutron dosimeter and detector. A thermal neutron device based on a portable D-D neutron generator has been designed and optimized by Monte Carlo method. The goal of optimal design is to maximize the thermal neutron flux at the output surface and improve the utilization efficiency of the neutron generator. The simulation of the reference device is experimentally verified. The contributions of different types and sizes of moderators, reflectors, and collimators to the thermal neutron intensity were simulated to obtain the optimum dimensions of the proposed design. Two parameters P and K were defined to describe the performance of device. It was determined that the final device uses high-density polyethylene (HDPE) as the moderator, nickel as the reflector, and HDPE as the collimator. The parameter K of the final device was 1.72. The P-value was 8 times that of the reference device. The thermal, epithermal + intermediate, fast, and total neutron flux on the output surface were 4.65, 2.52, 1.86, and 2.72 times of the reference device, respectively. The ratio of thermal to fast neutron flux ϕ th /ϕ fast increased from 0.52 to 1.30. The neutron irradiation device is supposed to be employed principally for neutron activation analysis, testing and calibration of neutron dosimeter and detector.

Verification of CENDL-3.2 Nuclear Data on VENUS-3 Shielding Benchmark by ARES Transport Code

The recently released CENDL-3.2 nuclear data library is deemed as an important achievement in the field of nuclear data research in China. To verify the applicability of the library to the shielding calculation of PWR and analyze the influence of multigroup cross-section parameters on the shielding calculation, ARES-MACXS module is used to process the MATXS format multigroup library based on CENDL-3.2 to generate multigroup working cross sections for PWR shielding calculation. VENUS-3 experimental facility has a clear and complete geometry. It is often used to test the ability of the advanced transport calculation method of calculating RPV fast neutron flux and to evaluate the accuracy of cross-section library. Different cross-section parameters are chosen for ARES to calculate VENUS-3 benchmark, and equivalent neutron flux of 58Ni(n,p)58Co, 115In(n,n′)115mIn and 27Al(n,α)24Na detectors is calculated according to the data provided by the benchmark report. The numerical results demonstrate that almost all the relative deviations between the calculated results and the experimental results are within 20%, which satisfies the requirement of shielding calculation. CENDL-3.2 is suitable for PWR shielding calculation. The comparison of various cross-section parameters results indicates that multigroup cross-section parameters have large effects on the transport calculation results.

State-of-the-art progress of gaseous radiochemical method for detecting of ionizing radiation

Abstract The article provides a review of the research results obtained during of more than 20 years concerning using the gaseous radiochemical method (GRCM) for detecting of ionizing radiation. This method based on threshold nuclear reactions with production of radioactive noble gas which does not interact with the materials of gaseous tract. The applications of GRCM in the diagnostics of neutrinos, neutrons, charged particles, thermonuclear plasma thermometry, and the study of the structure and dynamics of astrophysical objects, position-sensitive dosimetry of neutron targets with accelerator driving, spatial distribution of the fast neutron flux density in a nuclear reactor allowing the transformation of longitudinal coordinate of neutron flux distribution into a temporal distribution of the radiochemical gas decay counting rate (“barcode” semblance) and measurement of bombarding particles spectra are described. Experimental testing of the described technologies was made on the neutron target driven with the linear proton accelerator of Institute for Nuclear Research of Russian Academy of Sciences (INR RAS).

A new structure design to extend graphite assembly lifespan in small modular molten salt reactors

Dimensional change of the graphite assemblies due to irradiation is a key factor in limiting the service lifespan of the reactor core in small modular molten salt reactors. The graphite assembly structure, determining the fast neutron flux distribution and temperature distribution, has a notable impact on its dimensional change and hence its lifespan. In this article, we put forward an innovative solid hexagonal prism assembly (HPA) surrounded by fuel salt to prolong the lifespan of the graphite assemblies and compare it with the traditional hexagonal round channel assembly (RCA) with fuel salt flowing through its central round channel. Under the single-cell model, the calculation results prove that the neutron flux distribution and temperature distribution in the HPA are more uniform than those in the RCA. The HPA deforms more slowly than the RCA under the same operating conditions. The lifespan of the RCA, defined as when it expands back to its original size, is 7.5 years under the 100 MW/m3 power density of fuel salt, while the lifespan of the proposed HPA is 8.4 years. The HPAs also allow a further expansion in the radial direction due to its noncontact arrangement in the core. It will take 14.9 years for the HPA assemblies to contact each other, which displays a significant improvement in lifespan.

Investigation of irradiated metal of WWER-type reactor internals after 45 years of operation. Part 2. Calculated and experimental determination of the fast neutron fluence and damage dose

Fast neutron flux, fluence and damage dose have been determined when using experimentally measured specific activity of 54 Mn и 58 Co isotopes for metal of samples cut out from elements of pressure vessel internals of Novovoronezh NPP Unit No 3 (18Cr-10Ni-Ti steel, analog of AISI 321 steel). The results have been compared with the values calculated by KATRIN-2.5 computer code.

Verification and Validation of the Geant4 Monte Carlo Code Toolkit for DEMO TBR Evaluations

Verification and Validation analyses of the Monte Carlo code toolkit Geant4 were conducted with nuclear data from the JEFF-3.3 library to show the suitability for fusion applications. We present benchmarking against data of the engineering design oriented HCPB (Helium Cooled Pebble Bed) Tritium Breeder Module (TBM) mock-up experiment taken from the SINBAD compilation and nuclear analyses in comparison with MCNP of the DEMO reactor equipped with a HCPB breeding blanket. Missing functionalities for Geant4 were developed such as a track length detector with multiple energy bins and tally multiplication, reflective boundaries, and statistical checks in a post-processing step. The activation foil measurements in the TBM mock-up revealed Geant4 underestimates the uncollided and fast neutron flux increasingly with penetration depth, while being consistent with MCNP for the low energy neutron flux. Geant4 underestimates the experimental tritium activity in the breeder pellets by up to 2.7% more than MCNP, which is within the experimental uncertainty. The DEMO analyses yielded a 0.97% lower Tritium Breeding Ratio (TBR) of Geant4 than MCNP for a homogenized breeder blanket, and a 1.46% higher TBR of Geant4 than MCNP for a mostly heterogenic blanket. This good agreement with MCNP demonstrates Geant4’s suitability for DEMO TBR evaluations.

Neutron background measurement for rare event search experiments in the YangYang underground laboratory

Several experiments have been conducted in the YangYang Underground Laboratory in the Republic of Korea such as the search for dark matter and the search for neutrinoless double-beta decay, which require an extremely low background event rate due to the detector system and the environment. In underground experiments, neutrons have been identified as one of the background sources. The neutron flux in the experimental site needs to be determined to design a proper shielding system and for precise background estimation. We measured the neutron spectrum with a Bonner sphere spectrometer, with Helium-3 ($^{3}$He) proportional counters. The neutron flux at the underground laboratory was so low that the radioactive decays from the radioisotopes contained in the detector created a significant background interference to the neutron measurement. Using Monte Carlo simulations, the intrinsic $\alpha$ background distribution due to the radioactive isotopes in the detector materials, was estimated. The neutron count rate of each Bonner sphere was measured from the pulse height spectrum of the $^{3}$He proportional counter, after subtracting the $\alpha$ particle background. The neutron flux and the energy spectrum were determined using the unfolding technique. The total neutron flux measured was (4.46 $\pm$ 0.66) $\times$ $10^{-5}$ $\rm{cm^{-2} s^{-1}}$, and the thermal and fast neutron flux (in the range 1 to 10 MeV) were (1.44 $\pm$ 0.15) $\times$ 10$^{-5}$ $\rm{cm^{-2} s^{-1}}$ and (0.71 $\pm$ 0.10) $\times$ 10$^{-5}$ $\rm{cm^{-2} s^{-1}}$, respectively.

Influence of fast neutron irradiation on the phase composition and optical properties of homogeneous SiOx and composite Si–SiOx thin films

Layers and devices utilizing semiconductor nanocrystals have been the subjects of intensive research due to applications in opto- and microelectronic devices, solar cells, detectors, memories and in many more fields. We have shown previously that those nanocrystals in dielectric matrices undergo a substantial reformation during electron irradiation. The research of the interaction between semiconductor nanoclusters and irradiation is important for both the intentional modification of the structures and for understanding the stability of those devices under harsh, radiative conditions (e.g. space, nuclear, medical diagnosis, or similar applications). In the present research, we investigated the influence of neutron irradiation on substoichiometric silicon oxide. We investigated both homogeneous case and inhomogeneous case of matrices with silicon nanoclusters. We found that a fast neutron flux of 5.5 × 1013 neutrons/cm2 s and a fluence of 3.96 × 1017 neutrons/cm2 induce phase separation in the homogeneous films, whereas it decreases the volume fraction of the amorphous silicon phase caused by the reducing size of amorphous nanoclusters in the inhomogeneous films.

Measurement of muon-induced high-energy neutrons from rock in an underground Gd-doped water detector

We present a measurement of the rate of correlated neutron captures in the WATCHBOY detector, deployed at a depth of approximately 390 meters water equivalent (m.w.e.) in the Kimballton Underground Research Facility (KURF). WATCHBOY consists of a cylindrical 2 ton water target doped with 0.1% gadolinium, surrounded by a 40 ton undoped water hermetic shield. We present a comparison of our results with the expected rate of correlated neutron captures arising from high-energy neutrons incident on the outside of the WATCHBOY shield, predicted by a hybrid FLUKA/GEANT4-based simulation. The incident neutron energy distribution used in the simulation was measured by a fast neutron spectrometer, the 1.8-ton Multiplicity and Recoil Spectrometer (MARS) detector, at the same depth. We find that the measured detection rate of two correlated neutrons is consistent with that predicted by simulation. The result lends additional confidence in the detection technique used by MARS, and therefore in the MARS spectra as measured at three different depths. Confirmation of the fast neutron flux and spectrum is important as it helps validate the scaling models used to predict the fast neutron fluxes at different overburdens.

Measurement of Høgdahl neutron flux parameters in the SM1 subcritical reactor of the University of Pavia

• Subcritical neutron reactors provide low but stable neutron flux. • Høgdahl conventional flux characteristics are measured by bare irradiation method. • Flux parameters trends intra- and inter-channels are evaluated. Neutron flux of the thermal subcritical multiplication complex located at the nuclear pole of the University of Pavia was investigated to acknowledge the flux characteristics according to the Høgdahl convention in order to predict the expected activity of radionuclides while performing neutron activations. Flux parameters were evaluated by means of bare multi-monitor methods by preparing samples containing Au, Cr, Zn and Ni that were used to measure the epithermal shape correction factor, the conventional thermal to epithermal ratio and the thermal, epithermal and fast neutron fluxes in two irradiation channels of the facility. Despite high uncertainties affecting epithermal shape correction factors, thermal to epithermal ratios was measured with uncertainties ranging between tens and a few percent. In addition, conventional neutron fluxes were individually determined with a few percent level uncertainties, which are also suitable in applications requiring a known and/or ultra-stable neutron exposure.

Preparation by the double extraction process with preliminary neutron irradiation of yttria or calcia stabilised cubic zirconium dioxide microspheres

Abstract A modern approach to nuclear energy involves reprocessing like transmutations of spent nuclear fuel products to reduce their radiotoxicity and time needed for their storage. For this purpose, they are immobilized in inert matrices made of zirconia and can be burned in fast neutron reactor or Accelerator Driven System. These matrices in spherical form can be obtained by sol-gel process. The paper presents a method of microspheres fabrication based on the combined Complex Sol-Gel Process and double extraction process consisting in the preparation of zirconium-ascorbate sol and simultaneous extraction of water and nitrates. The procedure allows obtaining gel microspheres with a diameter of 50 μm, which after heat treatment are processed into the final product. The synthesis of zirconia microspheres with Yttrium by internal gelation process is well known for over a decade now. However, the explanation and characterization of synthesis of such material by extraction of water process is rarely found. Parameters such as: pH, viscosity, shape, sphericity and crystal structure have been determined for synthesized products and semi-products. In addition, preliminary research consisting in irradiation of the obtained materials in fast and thermal neutron flux was carried out. The obtained results are presented and described in this work.