Design and verification of a multi-layer single-sphere neutron spectrometer with water as the moderator

Measuring the neutron energy spectrum is important in nuclear radiation detection and protection. Common neutron spectrometers include the Bonner sphere spectrometer (BSS), time-of-flight neutron spectrometer, and plastic scintillation detector. Among them, the BSS is the most widely used for its wide measurement range and simple operation. A BSS usually occupies a large space because it contains several independent spheres working at the same time, leading to poor consistency. This paper proposes a multi-layer single-sphere spectrometer using water as the moderator. The spectrometer includes a multi-layered sphere that can be filled with water or air and a ^3He proportional counter placed in the center of the sphere. To verify the feasibility of this design, we use Geant4 to calculate the moderating ability of water and the response functions of the multi-layer single-sphere spectrometer. Additionally, several standard neutron energy spectra (from IAEA) are used to test the response characteristics of this spectrometer through simulation. The simulation results prove the feasibility of the design. This research provides a theoretical basis for a multi-layer single-sphere neutron spectrometer with water as the moderator.

Neutron energy spectrum measurement using CLYC7-based compact neutron emission spectrometer in the Large Helical Device

Tangential compact neutron emission spectrometer (CNES) based on the Cs2LiYCl6:Ce with 7Li-enrichment (CLYC7) scintillator is newly installed in the Large Helical Device (LHD). Measurement of neutron energy spectrum was performed using CNES in tangential neutral beam (NB) heated deuterium plasma discharges. The Doppler shift of neutron energy according to the direction of tangential NB injection has been obtained. When the fast ions moving away from the CNES, lower shifted neutron energy is obtained, whereas the upper shifted neutron energy is obtained when the fast ions moving toward the CNES. The obtained neutron energy is almost consistent with the virgin deuterium-deuterium neutron energy evaluated by the simple two-body kinematic calculation.

Iterative diffraction pattern retrieval from a single focal construct geometry image

Understanding the crystal structure of materials under extreme conditions of pressure and temperature has been revolutionized by major advances in laser-driven dynamic compression and in situ X-ray diffraction (XRD) technology. Instead of the well known Debye–Scherrer configuration, the focal construct geometry (FCG) was introduced to produce high-intensity diffraction data from laser-based in situ XRD experiments without increasing the amount of laser energy, but the resulting reflections suffered from profoundly asymmetrical broadening, leading to inaccuracy in determination of the crystal structure. Inspired by fast-neutron energy spectrum measurements, proposed here is an iterative retrieval method for recovering diffraction data from a single FCG image. This iterative algorithm restores both the peak shape and relative intensity with rapid convergence and requires no prior knowledge about the expected diffraction pattern, allowing the FCG to increase the in situ XRD intensity while simultaneously preserving the angular resolution. The feasibility and validity of the method are shown by successful recovery of the diffraction pattern from both a single simulated FCG image and a single laser-based nanosecond XRD measurement.

Calculation of neutronic characteristics in different reflector materials with a 15-MWt reactor core using VVR-KN fuel type

VVR-KN is one of the low enriched fuel types intended for a research reactor of a newCentre for Nuclear Energy Science and Technology (CNEST) of Viet Nam. As a part of design orientation for the new research reactor, the calculations of neutronic characteristics in a reactor core reflector using different materials were carried out. The investigated core configuration is a 15-MWt power loaded with VVR-KN fuel assemblies and surrounded by a reflector using beryllium, heavy water or graphite respectively. MCNP5 code together with up-to-date nuclear data libraries were used for these calculations. This paper presents the calculation results of neutron energy spectrum, neutron spatial distribution in the reflector using the above-mentioned materials. Besides, neutronic characteristics calculated for silicon doping irradiation holes in the reflector are also presented and the utilization capabilities of different reflector materials are discussed.

Space to Air High-Altitude Region Adjoint Neutron Transport

Neutrons from an atmospheric nuclear explosion can be detected by sensors in orbit. Current tools for characterizing the neutron energy spectrum assume a known source and use forward transport to recreate the detector response. In realistic scenarios the true source is unknown, making this an inefficient, iterative approach. In contrast, the adjoint approach directly solves for the source spectrum, enabling source reconstruction. The time–energy fluence at the satellite and adjoint transport equation allow a Monte Carlo method to characterize the neutron source’s energy spectrum directly in a new model: the Space to High-Altitude Region Adjoint (SAHARA) model. A new adjoint source event estimator was developed in SAHARA to find feasible solutions to the neutron transport problem given the constraints of the adjoint environment. This work explores SAHARA’s development and performance for mono-energetic and continuous neutron energy sources. In general, the identified spectra were shifted towards energies approximately 5% lower than the true source spectra, but SAHARA was able to capture the correct spectral shapes. Continuous energy sources, including real-world sources Fat Man and Little Boy, resulted in identifiable spectra that could have been produced by the same distribution as the true sources as demonstrated by two-dimensional (2D) Kolmogorov–Smirnov tests.

Neutron Beam Characterization at Neutron Radiography (NRAD) Reactor East Beam Following Reactor Modifications

The Neutron Radiography Reactor at Idaho National Laboratory (INL) has two beamlines extending radially outward from the east and north faces of the reactor core. The control rod withdrawal procedure has recently been altered, potentially changing power distribution of the reactor and thus the properties of the neutron beams, calling for characterization of the neutron beams. The characterization of the East Radiography Station involved experiments used to measure the following characteristics: Neutron flux, neutron beam uniformity, cadmium ratio, image quality, and the neutron energy spectrum. The ERS is a Category-I neutron radiography facility signifying it has the highest possible rank a radiography station can achieve. The thermal equivalent neutron flux was measured using gold foil activation and determined to be 9.61 × 106 ± 2.47 × 105 n/cm2-s with a relatively uniform profile across the image plane. The cadmium ratio measurement was performed using bare and cadmium-covered gold foils and measured to be 2.05 ± 2.9%, indicating large epithermal and fast neutron content in the beam. The neutron energy spectrum was measured using foil activation coupled with unfolding algorithms provided by the software package Unfolding with MAXED and GRAVEL (UMG). The Monte-Carlo N-Particle (MCNP6) transport code was used to assist with the unfolding process. UMG, MCNP6, and measured foil activities were used to determine a neutron energy spectrum which was implemented into the MCNP6 model of the east neutron beam to contribute to future studies.

Preliminary evaluation of the LASSO method for prediction of the relative power density distribution in mixed oxide (Pu,DU)O2 fuel pellets

Predicting the power distribution within nuclear fuel is essential for predicting reactor fuel performance, since power distributions can impact pellet temperature distributions and fission product transport and migration. Analytical expressions for radial power distribution in fuel pellets were sought using lattice physics calculations to generate data and a machine learning technique was applied to find representative expressions. Analytical approximations can be useful in nuclear fuel performance codes, such as ELESTRES/ELOCA for providing very rapid predictions of power distributions with reduced computational effort and memory requirements, relative to using an embedded or coupled neutron transport / burnup reactor physics code. Radial power distributions were calculated a priori using lattice physics codes to model mixed oxide (MOX) 37-element fuel bundles in pressure tube heavy water reactors (PT-HWRs). Such advanced fuels are of interest for future fuel cycles. Several datasets were generated with different amounts of PuO2 and variable neutron energy spectrum. Results of preliminary studies with the Least Absolute Shrinkage and Selection Operator (LASSO) regression machine learning method have obtained analytical fitting functions with a mean maximum relative error (MRE) of 0.056 and a maximum MRE of 0.152 on the test set. However, using LASSO to estimate the coefficients of a physically-motivated modified Bessel plus an exponential function, results in a lower MRE (mean MRE 0.041 and maximum MRE 0.11) on the same test set. Further potential improvements in both the curve fit and the machine learning methods are discussed.