High-intensity Neutron Flux Research Articles

The potential of the ILC beam dump for high-intensity and large-area irradiation field with atmospheric-like neutrons and muons

We evaluate the neutron and muon fluxes produced in the ILC beam dumps by Monte Carlo simulations and discuss their potential use in irradiation fields. We show that the beam dumps can provide high-intensity neutron and muon fluxes with spectra quite similar to secondary cosmic rays, which are suitable for soft error studies. The beam dumps deliver neutrons about 1011 times the secondary cosmic rays on spaces perpendicular to the beam axis and muons 108 times downstream of the beam dumps in the initial phase of the ILC. Large-area irradiation of 1m2 or more is possible. Differences in the energy distribution of muons in electron and positron beam dumps are also discussed for particle physics experiments.

Status and future developments of the Linear IFMIF Prototype Accelerator (LIPAc)

LIPAc is the Linear IFMIF Prototype Accelerator developed within the framework of the IFMIF project under the Broader Approach (BA) agreement signed between EURATOM and the Japanese Government in 2007. The IFMIF accelerator aims to provide an accelerator-based D-Li neutron source to produce high intensity neutron fluxes with appropriate energy spectrum in order to characterize materials envisioned for future fusion reactors. Because the IFMIF accelerator has to reach unprecedented performances, the feasibility is being tested through the design, manufacturing, installation, commissioning and testing activities of a 1:1-scale prototype accelerator, namely LIPAc, from the injector to the first cryomodule together with the High Energy Beam Transport line and the High Power Beam Dump. After outstanding results obtained in 2019, the LIPAc project has entered 2020 in the preparation of the third commissioning stage, i.e., validation in continuous-wave mode of the complete accelerator up to 5 MeV with its final beam dump. The validation until the nominal energy of 9 MeV will be made after the completion of cryomodule assembly. After a brief overview of the goals already achieved in the framework of the IFMIF/EVEDA program, this paper will present a synthesis of the results that have been obtained so far with the LIPAc accelerator as well as the future developments planned beyond 2020.

Experimental determination of the induced activity in activation detectors of a complex geometric shape

The paper presents the results of experimental determination of the induced activity in copper-/aluminum-based activation detectors when irradiated with neutrons with energies of about 14 MeV. The activation detectors were square-shaped metal plates with a thickness from 1.0 to 1.5 mm and a side size about 5.0 cm. These dimensions significantly exceed those of the detectors that are used in the research of high-intensity neutron fluxes. The detectors described in this work can be used for studying low-intensity neutron fluxes (with a flux density of up to 106n/cm2∙s). It is shown that, when working with such detectors, it is possible to apply the usual methods for calculating the induced activity in thin activation detectors, with corrections that take into account the emerging features of the ‘neutron source - activation detector’ and ‘activation detector - secondary radiation detector’ geometries. The effects of absorption of primary and secondary radiation by the detector substance are also revealed.The Geant4 tools were used for calculating the geometric factors and theoretical induced activity. The study confirms the applicability of such activation detectors for solving the problem of determining the yield of neutrons with energies of about 14 MeV from a neutron generator target. The results of the experiments coincide, within the margin of error, with the results of simulations performed using the Geant4 tools.

Three-dimensional magneto convective flows in geometries relevant for DCLL blankets

Abstract The dual coolant lead lithium (DCLL) blanket is one of the liquid metal blanket concepts investigated within the framework of the EUROfusion Consortium. In the proposed design, the liquid metal serves as tritium breeder, neutron multiplier, and as coolant for the breeding zone. Magnetohydrodynamic (MHD) effects resulting from the interaction of the moving liquid metal with the magnetic field that confines the fusion plasma are coupled with thermal phenomena. Induced electromagnetic forces lead to significant modifications of the hydrodynamic flow and affect in turn heat and mass transfer. The presence of a high-intensity neutron flux in the liquid metal yields volumetric heating and associated buoyant forces that drive complex flow patterns in the blanket. The resulting convective motions cause strong thermal mixing, and an increment in the effective heat transfer coefficient that may increase heat losses from the liquid metal into the helium coolant. This could deteriorate the thermal efficiency of the blanket. Buoyant MHD phenomena may also result in reversed flow and closed recirculations where tritium accumulates and temperature locally increases. The aim of the present investigation is to address by means of 3D numerical simulations the main features of magneto-convective flows in a geometry typical for DCLL blankets, in which PbLi moves upwards in a poloidal duct that faces the first wall and, after making a U-turn, flows downwards at the back of the blanket module. The liquid metal enters and leaves the model geometry by flowing radially in rectangular channels. It is found that in the front duct the velocity is characterized by jets in the boundary layers along walls parallel to the magnetic field, as expected in case of electrically conducting ducts. In the rear channel, along the wall that separates the two poloidal ducts, a downward flow occurs with much higher velocities than in the rest of the channel. The influence of the magnetic field on these flow features is investigated.

The degradation of gamma-ray mass attenuation of UOX and MOX fuel with nuclear burnup

Nondestructive gamma-ray spectrometry of nuclear fuel is routinely performed in axial gamma scanning devices and more recently with gamma emission tomography. Following the irradiation of a fresh nuclear fuel with high intensity neutron flux in a nuclear reactor core, a great number of gamma-emitting radionuclides are created. These can be utilized for gamma spectrometric techniques. However, due to the high density and atomic number of the nuclear fuel, self-attenuation of gamma-rays is a challenge, which requires attenuation correction in order to perform accurate analysis of the source activity in the fuel.In this study, the degradation of the gamma-ray mass attenuation with burnup was investigated and, in addition, a predictive model was created by investigating the attenuation change at various gamma energies caused by the burnup of the nuclear fuel. This model is intended for use by spectrometry practitioners inspecting nuclear fuel. To this aim, the energy-dependent gamma-ray mass-attenuation coefficients were investigated as a function of burnup for UOX, and three MOX fuels having different initial Pu contents. The Serpent 2 reactor physics code was used to simulate the burnup history of the fuel pins. The nuclide inventory of the Serpent 2 output is combined with the NIST XCOM database to calculate the mass attenuation coefficients.The mass attenuation coefficient of the fuel was found to decrease with the fuel burnup, in the range of a few percent, depending on the burnup and gamma energy. Also, a theoretical burnup dependent swelling model was imposed on fuel density to see how linear attenuation coefficient of fuel material is changed. Furthermore, greater effect may be expected on the transmitted intensity, where a simulation study of a PWR assembly revealed that the contribution from the inner rods in a scanned fuel assembly increased by tens of percent compared to the one with non-irradiated fresh fuels, when shielded by the outer rods of the assembly. A sensitivity analysis was performed in order to test the effect of a number of geometrical and operational reactor parameters that were considered to potentially effect the mass attenuation coefficient. Finally, a simple-to-use predictive model was constructed providing the mass-attenuation coefficient [cm2/g] of fuel as a function of burnup [MWd/kgHM] and initial Pu content [wt%]. The resulting predictive model was optimized by using the nonlinear regression method.

Investigation of the chance structure of the fission cross section of 238U by neutrons in energy range from 14 to 18 Mev using the delayed neutrons

In the present work, delayed neutrons are used for the investigation of the chance structure of the fission cross section. For this, time parameters of delayed neutron precursors emitted from the fission of 238U by neutrons with the energies from 14 to 18 MeV have been measured. The T(d,n) reaction was used as a neutron source. The experimental data have been processed to account the effect of degradation of detector counting rate in high intensity neutron flux and the effect of concomitant neutron source D(d,n) caused by implanted deuterium ions in neutron target. The data obtained based on the time parameters of delayed neutron precursors have been used in the analysis of the chance structure of the fission cross section of the 238U.

New study of various target neutron yields from spallation reactions using a high-energy proton beam

The spallation target plays an important role in the construction of an accelerator-driven system. Its purpose is to generate a neutron flux produced by cascaded spallation reactions using heavy nuclei, the latter being bombarded by the high-intensity proton beam. In this study, we have examined several heavy materials, such as uranium, thorium, tungsten, tantalum, lead, bismuth and mercury. The aim of this is to optimise a high-intensity neutron flux to be useful in several fields of applications, such as medicine and transmutation of nuclear waste. In this paper, we have shown when the spallation target changes the neutron flux varies. For example, uranium and thorium two materials producing a very intense amount of neutrons followed by lead, tungsten, mercury and bismuth and lately tantalum. We found these results by the variation of the proton beams energy from 0.1 GeV to 3 GeV, then with the variation of the geometry. Finally, we validated this study with experimental and theoretical results.

Simulation of the Effect of Burnup of Materials of an Ionization-Chamber Radiator by the Monte Carlo Method

Using the Monte Carlo method, the authors have modeled the burnup of fissionable material of the radiator of an ionization chamber in a neutron flux. A study has been made of the change in the sensitivity of the ionization chamber with time because of the burnup of fissionable material of the radiator for low- and high-intensity neutron fluxes having different energies.

A proposal for an alternative use of prompt-Self Powered Neutron Detectors: Online spectral-deconvolution for monitoring high-intensity neutron flux in LFRs

Self-Powered Neutron Detectors (SPNDs) are currently used in reactors’ environment to sense the magnitude of neutron-fluxes, usually for spatial-distribution mapping of the fuel region as to optimize burn-up strategies. During the development of tailored instrumentation for Lead-cooled Fast Reactors, the possibility to perform online spectral deconvolution of fast neutron-fluxes was recognized. Seven geometrically similar SPNDs with different neutron-sensitive materials, have been characterized by the Monte Carlo transport code MCNPX. Thanks to a database of spectral sensitivities vs. neutrons’ energy, a mathematical deconvolution process from 7 electric current values virtually measured by SPNDs was started, retrieving spectral information, in terms of 7 energy windows, of the neutron-flux the detectors are subjected by. This paper describes the procedure that led to those results, prefiguring future development to improve proposed method.

Reducing the Peak-to-Average-Power-Ratio in Fusion Blankets

Breeding blankets with integrated first wall are one of the most critical components of nuclear fusion reactors. Blankets breeding zones are characterized by steep nuclear heating gradients due to the exothermic nuclear reaction 6Li(n, )T and the high intensity neutron flux in the proximity of the first wall. Non-uniformity in nuclear heating can generate sharp temperature gradients that deeply affect material properties. This conceptual study explores an original way to flatten nuclear heating profiles by proposing a blanket characterized by layers of different 6Li enrichment in the breeder region while maximizing Tritium Breeding Ratio (TBR) and power generation. Two types of fusion blanket are studied: (1) Helium Cooled Ceramic Reflector (HCCR) and (2) Dual Coolant Lead Lithium (DCLL). For HCCR, it is found in the optimal design case, that the power peak-to-average can be reduced by 47.85%, 42.45% and 54.13% in the front, middle and back channel respectively when compared to the reference design. On the other side, we found that this method of profile flattening is not appealing for DCLL, under the geometrical configuration and material selection in this particular blanket design, since most of nuclear heating is caused by photon heat deposition.

High-flux PGAA for milligram-weight samples

With the high-intensity cold neutron flux available at the Prompt Gamma Activation Analysis (PGAA) instrument of the research reactor FRM II at the Heinz Maier-Leibnitz Zentrum (MLZ), samples with a weight of 1 mg or even less can be investigated for their elemental compositions using the (n,γ) capture reaction. In such cases, the typical sample packing material for PGAA experiments made of 25 μm thick PTFE foil (ca. 80 mg) can be orders of magnitude more massive than the sample weight itself. Proper choice of the packing material and measuring conditions are then of the highest importance [1].

An integrated model for materials in a fusion power plant: transmutation, gas production, and helium embrittlement under neutron irradiation

The high-energy, high-intensity neutron fluxes produced by the fusion plasma will have a significant life-limiting impact on reactor components in both experimental and commercial fusion devices. As well as producing defects, the neutrons bombarding the materials initiate nuclear reactions, leading to transmutation of the elemental atoms. Products of many of these reactions are gases, particularly helium, which can cause swelling and embrittlement of materials.This paper integrates several different computational techniques to produce a comprehensive picture of the response of materials to neutron irradiation, enabling the assessment of structural integrity of components in a fusion power plant. Neutron-transport calculations for a model of the next-step fusion device DEMO reveal the variation in exposure conditions in different components of the vessel, while inventory calculations quantify the associated implications for transmutation and gas production. The helium production rates are then used, in conjunction with a simple model for He-induced grain-boundary embrittlement based on electronic-structure density functional theory calculations, to estimate the timescales for susceptibility to grain-boundary failure in different fusion-relevant materials. There is wide variation in the predicted grain-boundary-failure lifetimes as a function of both microstructure and chemical composition, with some conservative predictions indicating much less than the required lifetime for components in a fusion power plant.

Effects of high intensity neutron flux on in-situ spin-exchange optical pumping of 3He

Prior studies revealed neutron beam-induced alkali-metal spin relaxation in 3He cells polarized in-situ by spin-exchange optical pumping. These effects are minor for neutron beams of low to moderate intensity, but become important at the highest neutron flux levels available in the full spectrum beams of high flux neutron sources. It was found that the relaxation consisted of a fast and a slow component, but the origin of neither is understood. This work further explored the mechanisms of this effect by measuring the magnitude and time dependence of alkali-metal spin-relaxation rates as a function of N2 and 3He composition. The experiments were performed on a thermal neutron beam line at the Institut Laue-Langevin with a maximum neutron flux of 2.4 × 109 cm−2 s−1.

RETRACTED: Triple-GEM detectors for electron, proton and neutron beam diagnostics

Different detectors based on triple GEM technology, realized in Frascati during the last years, are presented. They have been used for the luminosity measurements at the DAFNE�-Factory; for the beam position monitor at the BTF beam-test facility in Frascati, for the UA9 experiment at the CERN H8 beam-line, and in the SPS circulating beam; and for the high-intensity neutron flux measurements at FNG and Frascati Tokamak Upgrade. We also report on the design, construction and test of a GEM compact time projection chamber (TPC), for beam tracking. A description of the detector construction and assembly, together with the results achieved during tests at BTF and CERN, is also given.

First neutron production utilizing J-PARC pulsed spallation neutron source JSNS and neutronic performance demonstrated

The J-PARC 1 MW pulsed spallation neutron source JSNS was successfully launched on 30th of May 2008. To demonstrate the unique features of the moderator design and the neutronic performance of JSNS the neutron spectral intensity, absolute neutron flux, and time structure of the neutron pulse shapes etc. have been measured using several neutron instruments since then. The measured energy spectra clearly revealed the feature of the para-hydrogen, as expected when designing the moderator. The measured neutron flux below 0.4 eV agreed with the corresponding design value within ±20%, thus suggesting that the JSNS design calculations to have been reliable. World-class high-resolution diffraction data could be recorded due to the suitability of design of the moderators and the instruments. Another world-class high-intensity neutron flux was also capable of being demonstrated due to the unique design of the large cylindrical coupled moderator.

Design of the 50 kW neutron converter for SPIRAL2 facility

SPIRAL2 is a facility for the study of fundamental nuclear physics and multidisciplinary research. SPIRAL2 represents a major advance for research on exotic nuclei. The radioactive ion beam (RIB) production system is comprised of a neutron converter, a target and an ion source. This paper is dedicated to the designing of the 50 kW neutron converter for the SPIRAL2 facility. Among the different variants of the neutron converter, the one based on a rotating solid disk seems quite attractive due to its safety, ease in production and relatively low cost. Dense graphite used as the converter's material allows the production of high-intensity neutron flux and, at the same time, the heat removal from the converter by means of radiation cooling. Thermo-mechanical simulations performed in order to determine the basic geometry and physical characteristics of the neutron production target for SPIRAL2 facility, to define the appropriate beam power distribution, and to predict the target behaviour under the deuteron beam of nominal parameters (40 MeV, 1.2 mA, 50 kW) are presented. To study the main physical and mechanical properties and serviceability under operating conditions, several kinds of graphite have been analyzed and tested. The paper reports the results of such measurements. Radiation damage is the most important issue for the application of graphite as neutron converter. It is well known that the thermal conductivity of the neutron-irradiated graphite is reduced by a factor of 10 from the initial value after irradiation. Difference in volume expansions between the matrix and the fiber results in serious damage of neutron-irradiated C/C composites. Calculations showed that at high temperature the effect of neutron radiation is not so critical and that the change in thermal conductivity does not prevent the use of graphite as neutron converter.

Performance measurement of the scintillator with optical fiber detector for boron neutron capture therapy

The thermal neutron flux can be easily measured in real time by using the scintillator with optical fiber (SOF) detector. However the irradiation damage under high-intensity neutron flux causes deterioration of the SOF detector due to radiation damage to the plastic scintillator in which 6LiF is blended. After irradiating the SOF detector for 4 h (thermal neutron fluence is approximately 2.0×10 13 neutrons/cm 2), the sensitivity of the SOF detector decreased by 3.0%. After irradiating the SOF detector for 2 months (thermal neutron fluence approximately 6.4×10 14 neutrons/cm 2), the sensitivity was reduced to 42% of baseline. Supposing that the thermal neutron fluence is 2×10 12 neutrons/cm 2 on the surface of a patient in a BNCT treatment, the sensitivity of the SOF detector is reduced by approximately 0.3%. This report presents investigations on the deterioration of the SOF detector in irradiation experiments.

Accelerator-Based Target Design and Optimization Driven by High-Energy Electron Beams and Proton Beams for the Analysis of Neutron Generation Characteristics

Accelerator-based target design and optimization are presented in this paper as an approach for the analysis of neutron generation and characteristics. Electron-based targets and proton-based targets driven by high-energy accelerator beams are investigated. The target plays an important role in the external neutron sources in which the target was driven by high-energy accelerator beams to generate neutrons. The optimization of target design in this work is to obtain the maximum generation of neutrons out of targets considering target material and geometry, accelerator beam energy, and beam size. A three-dimensional particle detection methodology and a surface matrix arithmetic technique were used to determine the spatial distribution of the source particles (electron and proton) and the total neutron generation from the target outer surfaces. Neutron generation and characteristics were analyzed based on the optimized targets regarding neutron spectrum, average energy, and average flux. Monte Carlo calculations were performed by using MCNPX to estimate the particle interaction inside the target and to calculate the neutrons escaping out of the target surfaces.Results in this work indicated that a high-energy (1-GeV) electron accelerator beam is capable of producing high-intensity neutron flux at the range of 1.60 × 1013 n/cm2·s of 1-mA electron. Compared to an electron accelerator beam, a proton beam (1 GeV) generates higher-intensity neutron flux at the level of 4.83 × 1013 n/cm2·s of 1-mA proton. The neutron generation ratio (neutron per incident particle escaping from the target) was computed as 0.76 neutrons per electron and 38.8 neutrons per proton for the selected targets. In the electron accelerator-based target, neutron generation was mostly through photonuclear reactions (88%), followed by prompt fission (12%). Neutron production in the target of the proton accelerator-based target was mainly due to spallation reactions (40%) and prompt fissions (48%). The optimized size of the target for the electron accelerator-based target, in terms of the volume, was about 16 times smaller than that for the proton accelerator-based target. The estimated neutron energy distribution was much narrower, with the electron accelerator target ranging from 1.0 × 10−3 to 30 MeV. In the proton accelerator target, the neutron energies ranged between 1.0 × 10−5 MeV and 1 GeV.

Gas dynamic trap as high power 14 MeV neutron source

It is now widely recognized that further progress toward a commercial fusion reactor critically depends on the availability of low activated materials to be operated for many years in the fusion neutron environment without degradation of their properties. Therefore, high power neutron source (NS) for extensive material tests is in great demand. The realistic NS can be built on the basis of the gas dynamic trap (GDT) which is one of the simplest system for magnetic plasma confinement. GDT is an axisymmetric mirror machine of the Budker–Post type, but with a high mirror ratio ( R>10), and operated with warm and relatively dense plasma. Thus, due to frequent collisions, the plasma confined in a trap is very close to isotropic Maxwellian, and hence many instabilities are not excited and plasma behavior is similar to a classical one. It is proposed to use the oblique injection of neutral tritium and deuterium beams with an energy of the order of 100 keV into the GDT “warm” target plasma (electron temperature T e≥750 eV) to produce high intensity neutron flux. In this case, due to the ionization and charge-exchange collisions with plasma particles the energetic atoms will turn into ions confined in the trap. The axial density profile of these anisotropic energetic ions will be strongly inhomogeneous. Namely, the maximum of the fast ion density will be located in the vicinities of mirrors and the minimum—at the middle plane of the device. Thus, a strongly inhomogeneous fusion neutron flux, mostly generated in the fast deuterium–tritium (D–T) ion collisions will appear. GDT NS looks to be rather flexible system. It enables to construct the source stepwise, starting from rather moderate parameters. Increasing the testing zone length and neutral beam power one can increase both neutron flux and the flux density. Calculations show that even in the modest version of the NS with T e∼250 eV quite appreciable neutron flux density in the range of 0.2–0.3 MW/m 2 can be produced. At present, the maximum temperature about 130 eV is obtained in the experiments at the model GDT device at Budker INP. Further increase of the electron temperature in the device requires higher neutral beam power and now is considered to be straightforward.

The Mini-Inca Project: Experimental Study of the Transmutation of Actinides in High Intensity Neutron Fluxes

A new experimental installation has been recently commissioned at the High Flux Reactor of Institut Laue-Langevin (ILL) in Grenoble (France). It gives access to high intensity neutron spectra from pure thermal (5.6 1014 n/s/cm2) to epithermal (2 1015 n/s/cm2). Several of low mass (10 µg) mono-isotopic targets of actinide elements are in the process of being irradiated and analyzed by a number of techniques, from nuclear spectroscopy to off-line mass spectrometry and innovative double-deposit fission micro-chambers.In the present paper we will present the first experiments carried out at the thermal neutron spectrum installation with 242Pu, 241-243Am samples.