Energy Neutron Source Research Articles

The study and design of a deuteron drift tube linear accelerator for middle energy neutron source

The paper concerns a room-temperature cross-bar H-mode (CH) drift tube linac (DTL) with KONUS (Kombinierte Null Grad Struktur) [1,2] beam dynamics. To make the acceleration in DTL cell more efficient, we studied the correlation between transit time factor (TTF) and structural coefficients, first. Furthermore, we developed a new code with Python to demonstrate the longitudinal dynamics more clearly. The code computationally generates clusters, bunch centers, and emittance growth in a single figure. Thus, the stabilization region and cluster evolution at various negative phases can be studied. Based on the above studies, we designed a 162.5 MHz CH-DTL to accelerate 10 mA D+ from 2.11 MeV to 3.25 MeV in continuous-wave (CW) mode. The proposed CH-DTL is a part of the Middle Energy Neutron Source (MENS). The dynamics and RF design were iterated to make the gap voltage error lower than 1 %. The initial beam is assumed to come from a Radio Frequency Quadrupole accelerator (RFQ). The geometries of the CH-DTL are optimized by using CST. Multiparticle tracking from LEBT to RFQ is performed with TraceWin and the transmission efficiency in the CH-DTL is 100 %.

Composition design and preparation of lithium lead titanate (Li2PbxTi1-xO3, 0.1 < x < 0.9): A novel tritium breeding ceramic

In deuterium–tritium (D-T) fusion reactors, the design of tritium breeding materials plays very critical role in tritium self-sufficiency. Compared to materials like Li2O, Li2TiO3 has a weaker tritium breeding capability, because of the lower density of Li. For this reason, there is a need to further explore new lithium-containing ceramics. In this paper we have designed a new tritium-breeding Li2PbxTi1-xO3. Tritium production calculations are performed on a 2 m thickness spherical shells of Li2PbxTi1-xO3 with four types of 6Li abundance (7.5 %, 20 %, 50 %, 90 %) and six species of neutron source energy（0.025 eV, 1 MeV, 2.5 MeV, 2.7 MeV, 7.4 MeV and 14.1 MeV）located at the center. It is shown that such a modification may result in the tritium breeding ratio increase to 1.4 in the sphere. Based on simulations, the Li2Pb0.5Ti0.5O3 was synthesized via the sol–gel route, and the microscopic morphology of Li2Pb0.5Ti0.5O3 were briefly investigated. Upon analysis, it's found that lead is uniformly present in the ceramic material in the form of Li2PbO3. In comparison with Li2TiO3 ceramics, Li2PbxTi1-xO3 exhibits stronger tritium breeding capability.

Simulation study of a rhodium self-powered neutron detector for irradiation of nuclear fuel and material reactor

To accurately judge the irradiation test cycle of nuclear fuel and material in-pile irradiation, and provide precise design inputs for the post-irradiation performance evaluation of fuel and material, the NIT-Rh self-powered neutron detector has been designed and developed. The sensitivity characteristics of the detector have been simulated and verified. This study introduces the basic structure and working principle of the rhodium self-powered neutron detector for irradiation of nuclear fuel and materials in the reactor. Additionally, a new computational model has been constructed by the Monte Carlo method, and the current sensitivity of different sizes of emitters and insulators has been analyzed. The model considers a more realistic neutron source and β-electron energy spectrum, accounts for the space charge effect on the signal in the insulator, and compares the simulated sensitivity with the reactor test sensitivity. The results show that the new rhodium self-powered neutron detector calculation model is more accurate, and the relative deviation between the simulated and reactor test sensitivities is only 3.06 %. It can be used for the in-heap irradiation test and provides a method for the application of the irradiation of nuclear fuel and material reactors.

Physics design of 14 MeV neutron generator facility at the Institute for Plasma Research

A high energy and high yield neutron source is a prime requirement for technological studies related to fusion reactor development. It provides a high-energy neutron environment for small-scale fusion reactor components research and testing such as tritium breeding, shielding, plasma-facing materials, reaction cross-section data study for fusion materials, etc. Along with ITER participation, the Institute of Plasma Research, India is developing an accelerator-based 14 MeV neutron source with a yield of 1012 n s−1. The design of the source is based on the deuterium–tritium fusion reaction. The deuterium beam is accelerated and delivered to the tritium target to generate 14 MeV neutrons. The deuterium beam energy and tritium availability in the tritium target are the base parameters of the accelerator-based neutron source design. The paper gives the physics design of the neutron generator facility of the Institute for Plasma Research. It covers the requirements, design basis, and physics parameters of the neutron generator. As per the analytical results generator can produce more than 1 × 1012 n s−1 with a 110 keV D+ ion beam of 10 mA and a minimum 5 Ci tritium target. However, the detailed simulation with the more realistic conditions of deuteron ion interaction with the tritium titanium target shows that the desired results cannot be achieved with 110 keV. The safe limit of the ion energy should be 300 keV as per the simulation. At 300 keV ion energy and 20 mA current, it reaches 1.6 × 1012 n s−1. Moreover, it was found that to ensure sufficiently long operation time a tritium target of more than 20 Ci should be used. The scope of the neutron source is not limited to the fusion reactor research studies, it is extended to other areas such as medical radioisotopes research, semiconductor devices irradiations, and many more.

Design optimization of the closed-water activation loop at the JSI irradiation facility

A closed-water activation loop is being built at the research reactor TRIGA at the Jožef Stefan Institute, Slovenia, which will serve as a well-defined and stable high energy gamma-ray and neutron source. The main focus of this work is to analyse and optimise the main components of the water activation loop to achieve the highest activity at reasonably low pressure drop. Both neutronic and hydraulic aspects were considered. The results show that the desired configuration consists of a high effective water volume of the inner activation part, outer observation part and narrow transport pipes connecting both parts. The complex snail-shape design of the inner part systematically outperformed other simplified designs by more than a factor of two for the main water activated isotopes (16N, 17N and 19O). The activity values increase strongly with increasing flow rate, but reach saturation at about 0.5 l/s. On this basis, the snail configuration was selected as the main candidate for both the inner activation and outer observation part. To not exceed the pressure drop of 3 bar within the closed-water loop, the nominal water flow rate and the inner diameter of the transport pipes were set to 0.5 l/s and 1.2 cm, respectively.

ENDF/B-VIII.0-based fast neutron removal cross sections database in Z = 1 to 92 generated via multi-layered spherical geometry

A fast neutron mass removal cross sections (ΣR/ρ) database covering the elemental range of Z = 1 to 92 is generated for source neutron energy distributions based on 241Am–Be, 252Cf, and 235U. Monte Carlo simulations using ENDF/B-VIII.0 nuclear data were performed with the fast neutron removal theory's ideal spherical representation. The ENDF/B-VIII.0-based ΣR/ρ′s derived were compared with values from the Oak Ridge National Laboratory's Lid Tank Shielding Facility (LTSF) experiments, values from software Phy-X/PSD and MRCsC, values obtained by an empirical model in Zoller, and values derived using the older ENDF/B-VII.0 nuclear data library. Subsequently, the generated ΣR/ρ′s were evaluated for convergence under varying moderator thickness. Results indicate that the ΣR/ρ quantity for fission and (alpha, n) spectra is universal, with slight deviations for low Z elements. The ΣR/ρ values generated from ENDF/B-VIII.0 agreed with LTSF experiments with an average relative deviation of 8.58%. Moreover, using ENDF/B-VIII.0-based ΣR/ρ′s, the derived fast neutron effective removal cross sections (ΣR's) for several light compounds and dry concrete have better agreement on average with LTSF experiments compared with values derived from the available ΣR/ρ′s in the literature. Significant differences were found for ΣR/ρ′s that were generated with ENDF/B-VIII.0 as compared with the superseded ENDF/B-VII.0. Furthermore, ENDF/B-generated ΣR/ρ′s through spherical model showed the expected convergence towards singular values at moderator thicknesses practical for fast neutron removal theory applications. Owing to the relatively high unavoidable uncertainties of the LTSF experimental ΣR/ρ′s, the ENDF/B-VIII.0-based ΣR/ρ′s present a new and potentially favorable database in ΣR general calculations for homogeneous multi-element materials for point-kernel shielding applications.

Determination of biological radioprotective characteristics of some natural organic compounds for radiation shielding applications

Abstract Ionizing radiation is used in many fields in energy, medicine, and industrial applications. Those who are in these areas or cancer patients receiving radiotherapy are at risk for acute or long-term exposure to radiation damage due to these ionizing radiations. Non-toxic new agents are needed to protect intact tissue and cells. In this study, we aimed to determine the gamma and neutron radiation attenuation characteristics of seven different natural compounds (quercetin, menadione, naphthol, caffeine, quinine sulphate, cholesterol and riboflavin) to help users in radiation applications. Gamma radiation attenuation parameters such as the mean free path, mass attenuation coefficient, effective atom number, linear attenuation coefficient, and half-value layer were calculated theoretically with WinXCom software for the energy range 0.015–15 MeV. Fast neutron attenuation criteria, such as mean free path, half-value layer, effective removal cross-sections and transmission neutron number, were theoretically determined with Monte Carlo simulation codes (Geant4). Neutron absorption measurement experiments were also applied in addition to the theoretical results. The neutron radiation absorption capacities were determined for samples with an 241Am-Be 4.5 MeV energy neutron source and portatif-type Canberra brand BF3 gas neutron detector. Neutron attenuation parameters were compared with paraffin to determine the absorption capability of the samples. It was found that the dose 1.1094 (μSv/h) from the source was absorbed by the samples to the following extent: 31.76% (Quercetin), 21.85% (Menadione), 28.85% (Naphthol), 22.94% (Caffeine), 12.51% (Quinine sulphate), 40.44% (Cholesterol) and 20.94% (Riboflavin). From the results, it can be clearly seen that all these drug samples had a good neutron radiation attenuation capacity. This revealed that the examined samples had radiation absorption abilities. It was found that the cholesterol sample had an especially excellent absorption power for both neutron and gamma radiation. The samples investigated in this study could be used to develop radiation-protective drugs.

Energy Dependence of Atmospheric Neutron-Induced Failures in Silicon Carbide Power Devices

A time-of-flight measurement technique is used to determine energy-dependent neutron cross sections of silicon carbide power devices exposed to a beam that approximates atmospheric neutrons up to 600 MeV. The Irradiation of Chip Electronics (ICE)-II neutron beam at the Los Alamos Neutron Science Center (LANSCE) is used during these experiments. The measurements show that a wide range of neutron energies are likely to cause device failures close to the rated voltage, from neutron interactions with lattice atoms within the active device volume. The data indicate that the energy-dependent cross sections are relatively large, revealing a high probability for a few percent of the neutron&#x2013;lattice atom interactions to give rise to a failure when the device is biased close to its rated voltage. Additionally, knowledge of the neutron energies and associated cross sections giving rise to failures within these devices provide insights regarding the critical charges and the sensitive volumes of these devices as a function of neutron energy for single-event burnouts. Lastly, the energy-dependent cross sections provide a preliminary set of data to calculate a correlation factor between failure rates obtained using a broad energy neutron source, and the commonly used mono-energetic high-energy proton beams, utilized due to their wider availability for determining bias dependent failure-in-time curves.

Characterization and qualification of neutron radiation effects – Summary of Japan-USA Joint Projects for 40 years –

The Joint Projects under the Japan-USA Fusion Cooperation Program started in 1981 and has continued for more than 40 years. In the Joint Projects, although a wide range of fusion materials and engineering issues were covered, neutron radiation effects on fusion reactor materials have always been the major research emphases, and the neutron irradiation facilities in the US were jointly used by Japanese and US researchers. Japanese test facilities including neutron and charged particle irradiation facilities were complementarily used.The initial focus of the Joint Projects was on fundamental fusion neutron radiation effects and irradiation correlation. Systematic comparison of fission and fusion radiation effects in comparable damage levels and the effects of transmutation-induced helium were investigated. The collaboration was then focused on the effect of dynamic irradiation effects in variable conditions. In addition to the relatively fundamental studies, the Joint Projects contributed largely to development of candidate materials such as RAFM steels, vanadium alloys, SiC/SiC composites, and tungsten alloys, through a mechanism-oriented approach. The Joint Projects also covered issues specific to materials application to fusion blankets and plasma-facing components, including neutron radiation effects such as tritium retention and permeation of neutron-irradiated plasma-facing materials. Various irradiation technologies were developed and applied to the irradiation experiments, including those for in-situ testing.Considering that high energy neutron sources, such as A-FNS and IFMIF-DONES, now have high viability, the research supporting the neutron source programs is essential. The knowledge obtained through the Joint Projects is valuable and should be advanced for this purpose. To this end, it is of urgent necessity to launch an international scientific program accumulating knowledge of fusion neutron radiation effects, including their fundamental aspects.

Impact of neutron energy on asteroid deflection performance

In the future, a hazardous asteroid will find itself on a collision course with Earth. For asteroids of moderate size or larger, a nuclear device is one of humanity’s only technologies capable of mitigating this threat via deflection on a timescale of less than a decade. This work examined how the output neutron energy from a nuclear device standoff detonation affects the deflection of a notional asteroid that is 300 meters in diameter and composed of silicon dioxide at a bulk density of 1.855 g/cm3. 14.1 MeV and 1 MeV neutron energy sources were modeled in MCNP to quantify the energy deposition in the asteroid target. The asteroid’s irradiated region was discretized in angle by tracing the rays emanating from the point of detonation and in depth by considering the neutron mean-free-paths. This high-fidelity approach was shown to deviate from previous analytic approximations commonly used for asteroid energy deposition. 50 kt and 1 Mt neutron yields of the energy deposition mappings were imported into a hydrodynamic asteroid model in ALE3D to simulate the deflective response due to blow-off ejecta. Underexplored in literature, changing the neutron energy was found to have up to a 70% impact on deflection performance due to induced differences in the energy deposition profile and in the energy coupling efficiency. The magnitude of energy deposition accounted for most of the observed variation in the asteroid velocity change, making the coupling efficiency more significant than the spatial profile characteristics. These findings are vital for determining the optimal source neutron energy spectrum for asteroid deflection applications.

EFFECT OF FISSION SOURCE SPECTRUM ON MONTE CARLO CALCULATION OF EX-CORE QUANTITIES

The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) offers unique capabilities to combine highfidelity in-core radiation transport with temperature feedback using MPACT and CTF with a follow-on fixed source transport calculation using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides a fission source to Shift for the follow-on radiation transport calculation. In past VERA releases, MPACT passed a spatially dependent source without the energy distribution to Shift. Shift then assumed a235U Watt spectrum to sample the neutron source energies. There were concerns that, in cases with burned or mixed oxide (MOX) fuel near the periphery of the core, the assumption of a235U Watt spectrum for the source neutron energies would not be accurate for studying ex-core quantities of interest, such as pressure vessel fluence or detector response. Therefore, two additional options were implemented in VERA for Shift to sample neutron source energies: (1) a nuclide-dependent Watt spectra for235U,238U,239Pu, and241Pu, and (2) to use the standard 51-energy group MPACT spectrum. Results show that the 51-group MPACT spectrum is not suitable for ex-core calculations because the groups have been fine-tuned for in-core calculations. Differences in relative detector response due to235U and nuclide-dependent Watt spectra sampling schemes were negligible; however, the use of nuclide-dependent Watt spectra for vessel fluence calculations was found to be important for fuel cycles with burned and fresh fuel.

Neutron instrumentation research at the Low Energy Neutron Source

Particle accelerators have been used to produce neutrons at relatively small-scale facilities for many decades, but historically the vast majority of the work performed at such facilities has focus...

Monte Carlo simulation of neutron shielding performance of iron spherical shell

Neutron Shielding with different energy is one of the important issues in radiation safety. To investigate the neutron shielding performance of iron spherical shell, this paper built a simple physical model of iron spherical shell; the neutron transport processes in iron spherical shell with different thicknesses were numerically simulated by using Monte Carlo method. For isotropic neutron source with energy ranging from 0.1 keV to 14 MeV, the numbers of leakage neutrons and gamma rays from iron spherical shell with different thicknesses has been obtained. The simulation results show that both the energy of neutron source and the thickness of iron spherical shell shield have a significant influence on number of leakage neutrons and gamma rays. For iron spherical shell shield, the number of leakage neutrons decreases with the increase of thickness of iron spherical shell when the normalized number of leakage neutrons is less than 1 (i.e. when the number of neutrons absorbed by iron is few), but increases with the increase of thickness of iron spherical shell when the normalized number of leakage neutrons is greater than 1 (i.e. when the reaction of (n, 2n) occurs in neutron transportation process). For neutron source with energy of 2 MeV, the normalized number of leakage neutrons is less than 1. The normalized number of leakage gamma rays keeps the trend from increase to reduction with the increase of thickness of iron spherical shell shield; the ratio of number of leakage neutrons to leakage gamma rays decreases with the increase of thickness of iron spherical shell shield for a certain range of thickness.

Monte Carlo optimization simulation of neutron shielding performance of iron/polyethylene combined structure

This study aims to get the optimization neutron shielding design of iron/polyethylene combined shield structure. The neutron transmission coefficient with various energies for different thickness of iron and polyethylene combined shield structure were calculated by using Monte Carlo method. The simulation results show that the optimization effect of iron/polyethylene combined shield is not obvious when the neutron energy is low or the shield is thin, there is an optimal thickness ratio of iron to polyethylene adopted to get the best neutron shielding performance when the energy of neutron source is above 2 MeV and the total thickness of combined shielding structure is more than 20 cm. The optimal thickness ratio of iron to polyethylene increases with the increasing energy of neutron source; with the increasing of neutron source energy ranging from 4 MeV to 14 MeV, the optimal thickness ratio of iron to polyethylene trends from 0.11 to nearly 1.6.

Benchmarking Geant4 for spallation neutron source calculations

Geant4 is becoming increasingly used for radiation transport simulations of spallation neutron sources and related components. Historically, the code has seen little usage in this field and it is of general interest to investigate the suitability of Geant4 for such applications. For this purpose, we carried out Geant4 calculations based on simple spallation source geometries and also with the the European Spallation Source Technical Design Report target and moderator configuration. The results are compared to calculations performed with the Monte Carlo N- Particle extended code. The comparisons are carried out over the full spallation neutron source energy spectrum, from sub-eV energies up to thousands of MeV. Our preliminary results reveal that there is generally good agreement between the simulations using both codes. Additionally, we have also implemented a general weight-window generator for Geant4 based applications and present some results of the method applied to the ESS target model.

Materials and neutronic research at the Low Energy Neutron Source

In the decade since the Low Energy Neutron Source (LENS) at Indiana University Center for Exploration of Energy and Matter (CEEM) produced its first neutrons, the facility has made important contributions to the international neutron scattering community. LENS employs a 13MeV proton beam at up to 4kW beam power onto one of two Be targets to produce neutrons for research in fields ranging from radiation effects in electronics to studies of the structure of fluids confined in nanoporous materials. The neutron source design at the heart of LENS facilitates relatively rapid hands-on access to most of its components which provides a foundation for a research program in experimental neutronics and affords numerous opportunities for novel educational experiences. We describe in some detail a number of the unique capabilities of this facility.

Seeking to bridge Europe’s impending neutron gap

Europe’s primacy in neutron scattering depends on maintaining a network of local and international facilities.

Neutron activation analysis for cement elements using an IECF device as a high energy neutron source

In this study, Neutron Activation Analysis (NAA) for cement elements using an Inertial Electrostatic Confinement Fusion (IECF) facility as a high energy neutron source is investigated.

A method for using neutron elastic scatter to create a variable energy neutron beam from a nearly monoenergetic neutron source

Abstract This work describes preliminary investigation into the design of a compact, portable, variable energy neutron source. The proposed method uses elastic neutron scatter at specific angles to reduce the energy of deuterium–deuterium or deuterium–tritium (D–T) neutrons. The research focuses on D–T Monte Carlo simulations, both in idealized and more realistic scenarios. Systematic uncertainty of the method is also analyzed. The research showed promise, but highlighted the need for discrimination of multiply-scattered neutrons, either through a pulsed generator or associated particle imaging.

Spin echo small angle neutron scattering using a continuously pumped (3)He neutron polarisation analyser.

We present a new instrument for spin echo small angle neutron scattering (SESANS) developed at the Low Energy Neutron Source at Indiana University. A description of the various instrument components is given along with the performance of these components. At the heart of the instrument are a series of resistive coils to encode the neutron trajectory into the neutron polarisation. These are shown to work well over a broad range of neutron wavelengths. Neutron polarisation analysis is accomplished using a continuously operating neutron spin filter polarised by Rb spin-exchange optical pumping of (3)He. We describe the performance of the analyser along with a study of the (3)He polarisation stability and its implications for SESANS measurements. Scattering from silica Stöber particles is investigated and agrees with samples run on similar instruments.