Fusion Neutron Spectra Research Articles

Simulation of unfolding fast neutron spectrum from CR-39 track morphology

The aim of this study is to investigate the impact of various factors on neutron spectrum unfolding, including track parameters (major axis, minor axis and depth), polyethylene (PE) thickness, and etching time. The Monte Carlo code Geant4 and proton track parameter calculation program TRACK_P were applied to simulate the response matrix and track parameters spectrum. Four test spectra—252Cf spectrum, Gaussian broadened deuterium–deuterium fusion neutron spectrum, 241Am–Be spectrum, and Gaussian broadened deuterium–tritium fusion neutron spectrum—were studied using the maximum-likelihood expectation maximization algorithm for neutron spectrum unfolding. The unfolded spectra were in good agreement with the reference spectra. Results showed that the major and minor axes were more suitable than the track depth for neutron spectrum unfolding. For high-energy neutrons, a thick PE layer is required. Moreover, etching time does not significantly affect the unfolded spectra. This study explores the factors influencing the unfolding of fast neutron spectrum based on simulation calculations and preliminary experiments that can be useful for applying CR-39 to neutron spectrum measurements.

Relevance of Tritium Breeder Irradiation Testing in a Fusion Prototypic Neutron Source

A fusion prototypic neutron source (FPNS) will play a fundamental role in testing and developing fusion materials that can withstand the harsh fusion environment. A thorough assessment is conducted to determine whether blanket materials would benefit from FPNS testing. Because of the cross sections for the various blanket materials, lithium-containing materials can generally be tested adequately using existing fission irradiation capabilities instead of waiting for or relying solely on future FPNS experiments. However, non-lithium blanket materials, such as beryllium multipliers and flow channel inserts, should be tested under a fusion neutron spectrum.

Physics-based model of irradiation creep for ferritic materials under fusion energy operation conditions

Irradiation creep is known to be an important process for structural materials in nuclear environments, potentially leading to creep failure at temperatures where thermal creep is generally negligible. While there is a great deal of data for irradiation creep in steels and zirconium alloys in light water reactor conditions, much less is known for first wall materials under fusion energy conditions. Lacking suitable fusion neutron sources for detailed experimentation, modeling, and simulation can help bridge the dose-rate and spectral-effects gap and produce quantifiable expectations for creep deformation of first wall materials under standard fusion conditions. In this paper, we develop a comprehensive model for irradiation creep created from merging a crystal plasticity representation of the dislocation microstructure and a defect evolution simulator that accounts for the entire cluster dimensionality space. Both approaches are linked by way of a climb velocity that captures dislocation-biased defect absorption and a dislocation strengthening term that reflects the accumulation of defect clusters in the system. We carry out our study in Fe under first wall fusion reactor conditions, characterized by a fusion neutron spectrum with average recoil energies of 20 keV and a damage dose rate of ≈3×10−7 dpa/s at temperatures between 300 and 800 K.

РАСЧЕТ ЭКСПЕРИМЕНТА ПО ОБЛУЧЕНИЮ КОНСТРУКЦИОННЫХ МАТЕРИАЛОВ В ПОТОКЕ ТЕРМОЯДЕРНЫХ НЕЙТРОНОВ КОДОМ ВЫГОРАНИЯ BPSD

The calculation of activation inventories is one of the key aspects in nuclear plants simulation. Burnup calculations provide information not only about fuel actinides transmutation but also about impact of fuel/constructor materials/coolant impurities on the material. The inventory evolution also determines radiological response of a material by nuclides production/decay rates quantification. Results of the isotopic kinetics code BPSD for activity and afterheat calculation validation by means of direct comparison afterheat obtained by code BPSD with experimental data obtained due to steel SS-304 and steel SS-316 samples irradiation by fusion neutron spectra at the FNS facility and with results of FISPACT-II calculation are presented in this paper. BPSD code is intended for carrying of transmutation calculations of materials in fast neutron fluxes. Nuclides transformation chains are based on ROSFOND and ABBN-RF data. BPSD and FISPACT-II calculations results conform to each other. Differences in results could be explained by differences between transformation chains and by using different evaluation neutron data bases.

Study on neutron spectrum unfolding method of organic scintillation measurement based on iterative regularization

An iterative regularization method is proposed for neutron spectrum unfolding of the liquid organic scintillator EJ-301. The Arnoldi iteration is used to generate Krylov subspaces. GMRES (Generalized Minimal Residual) combined with Tikhonov regularization is used for neutron spectrum updating. The restart method is applied to speed up the convergence as well as non-negativity constraints. The Monte Carlo code Geant4 is applied to simulate the response functions and PHSs (pulse height spectrums) of the neutron spectrums for unfolding. Four typical neutron spectrums are studied, such as the mono-energetic DD fusion neutron spectrum, 241Am-Be neutron spectrum, 252Cf neutron spectrum and Gaussian broadened DT fusion neutron spectrum. Comparison study between the proposed method and the existing methods is also carried out, such as GRAVEL, MLEM, and direct Tikhonov regularization. Unfolding results with acceptable accuracies are obtained with the proposed method, and the proposed method shows better adaptabilities than the other three methods with both the ideal and simulated PHSs in most cases.

Calibration study for a “fission electron-collection” neutron detector

Abstract The “fission electron-collection” neutron detector (FECND) is a fission-based neutron detector with flat and fast responses, and is a promising choice for the diagnostics of fusion neutron flux and spectrum. Due to its low sensitivity, the detector output is weak and is disturbed easily in calibration experiments. The backgrounds in the calibration primarily come from scattered neutrons and the γ rays induced by neutron inelastic scattering. A symmetrical structure of FECND was proposed to suppress the output of γ rays, and its effectiveness was validated by Monte Carlo simulations. A method was presented to eliminate the impact of scattered neutrons by utilizing the solid angle effect and the energy difference between source neutrons and scattered neutrons. Through calibration experiments and analyses, a D-T sensitivity of (8.6 ± 0.2)×10−20 C·cm2 was obtained for the FECND.

Study of transmutation, gas production, and displacement damage in chromium for fusion neutron spectrum

In the present work, gas production, transmutation and displacement damage in chromium are studied for fusion neutron spectra. The nuclear cross-section data are calculated with the TALYS-1.8 code. Transmutated isotopes in chromium, including the radioactive ones, are identified for typical fusion reactor neutron spectrum. Gas production per atom (He and H production) are also predicted for typical fusion neutron spectrum. Molecular dynamics (MD) simulations of damage cascade caused by the self-recoil in chromium are carried out for up to 200 keV damage energies using the LAMMPS code. Time evolution of interstitials and vacancies are studied and discussed. Based on the performed MD simulations, constant parameters of the athermal-recombination corrected dpa (arc-dpa) method are calibrated. Displacement damage cross-section of chromium are calculated using the Norgett-Robinson and Torrens (NRT) and arc-dpa approaches and later used to predict the values of displacement per atoms (dpa) in chromium for typical fusion reactor neutron environment.

Correction and verification of HL-2A Tokamak Bonner sphere spectrometer in monoenergetic neutron fields from 100 keV to 5 MeV

A real-time Bonner sphere spectrometer (BSS) has been developed for spectral neutron measurements with the HL-2A Tokamak. To correct and verify the accuracy of the neutron spectrum from the BSS, the BSS system was calibrated using monoenergetic neutron beams in the energy range of 100 keV–5 MeV. The response function of the BSS was corrected based on the calibration results, and the corrected BSS system was verified by unfolding monoenergetic neutron spectra. Fusion neutron spectra on the HL-2A have been obtained from the calibrated BSS system for the first time.

Study on unfolding method of neutron spectrum of BSS (Bonner Sphere Spectrometer) based on compressed sensing

The Bonner Sphere Spectrometer (BSS) is widely used for neutron spectrum measurement for its broad energy sensitivity (from thermal to GeV). Information of neutron spectrum cannot be directly obtained and must be unfolded from the measurement readings. In this paper, a novel method of unfolding neutron spectrum from BSS measurement based on compressive sensing (CS) is proposed. Sparse representation based on learning dictionary is employed. The orthogonal matching pursuit (OMP) algorithm and the K-SVD algorithm are applied for sparse coding and dictionary update. The SL0 algorithm is used for reconstruction. Computational experiment is carried out, in which 5 types of the neutron spectrum are tested, including the D–T fusion neutron spectrum, Watt fission spectrum, Maxwell fission spectrum, the evaporation spectrum and 241Am-Be neutron spectrum taken from ISO standard 8529-1. The computational experiment reveals that: firstly, with appropriate sparse representation, neutron spectrum can be unfolded with high accuracy; secondly, the unfolding method based on CS is very suitable for solving the highly under determined problem. For instance, neutron spectrum with 300 energy groups is unfolded with 7 equations.

Three-dimensional modeling of the neutron spectrum to infer plasma conditions in cryogenic inertial confinement fusion implosions

Neutron-based diagnostics are typically used to infer compressed core conditions such as areal density and ion temperature in deuterium–tritium (D–T) inertial confinement fusion (ICF) implosions. Asymmetries in the observed neutron-related quantities are important to understanding failure modes in these implosions. Neutrons from fusion reactions and their subsequent interactions including elastic scattering and neutron-induced deuteron breakup reactions are tracked to create spectra. It is shown that background subtraction is important for inferring areal density from backscattered neutrons and is less important for the forward-scattered neutrons. A three-dimensional hydrodynamic simulation of a cryogenic implosion on the OMEGA Laser System [Boehly et al., Opt. Commun. 133, 495 (1997)] using the hydrodynamic code HYDRA [Marinak et al., Phys. Plasmas 8, 2275 (2001)] is post-processed using the tracking code IRIS3D. It is shown that different parts of the neutron spectrum from the view can be mapped into different regions of the implosion, enabling an inference of an areal-density map. It is also shown that the average areal-density and an areal-density map of the compressed target can be reconstructed with a finite number of detectors placed around the target chamber. Ion temperatures are inferred from the width of the D–D and D–T fusion neutron spectra. Backgrounds can significantly alter the inferred ion temperatures from the D–D reaction, whereas they insignificantly influence the inferred D–T ion temperatures for the areal densities typical of OMEGA implosions. Asymmetries resulting in fluid flow in the core are shown to influence the absolute inferred ion temperatures from both reactions, although relative inferred values continue to reflect the underlying asymmetry pattern. The work presented here is part of the wide range of the first set of studies performed with IRIS3D. This code will continue to be used for post-processing detailed hydrodynamic simulations and interpreting observed neutron spectra in ICF implosions.

Fast heating of fuel assembled in a spherical deuterated polystyrene shell target by counter-irradiating tailored laser pulses delivered by a HAMA 1 Hz ICF driver

Fast heating is a method of heating an assembled high-density plasma into a hot state by irradiating it with short-duration (sub-picosecond), high-intensity ( W ) laser pulses before the plasma expands and dissolves hydrodynamically. In this paper, we present detailed experimental results of fast heating fuel assembled in a spherical deuterated polystyrene shell target of 500 m diameter and 7 μm thickness with counterbeam illumination by using a HAMA 1 Hz, 5.9 J inertial confinement fusion laser driver with pulse tailoring. These tailored pulses contain three pulses in sequence: a ‘foot’ pulse of 2.4 J/25 ns, a ‘spike’ pulse of 0.5 J/300 ps and a ‘heater’ pulse of 0.4 J/110 fs; these pulses are designed to assemble the fuel and heat it. By varying the energy of the foot pulse, we find that fast heating the fuel is achieved only if the fuel is weakly ablated by the foot pulse and then shock-assembled by the spike pulse into the target centre so that the heater pulse can access the fuel with a focal intensity greater than W . Without a foot pulse, the heater pulse contributes to assembling the fuel. For higher foot-pulse energies, the heater pulse drives a hydrodynamic motion with speeds of the order 107 cm with intensities of the order W , resulting in re-assembling and additional heating of the pre-assembled fuel. Once a shock-assembled core is achieved at the target centre, we succeed qualitatively in fast heating the core for shots in sequence with variations of laser energy within 18%. The coupling efficiency from the heating laser to the core is inferred to be % in total: % for the ionized bulk electrons and % for the bulk ions. The fusion neutron spectrum detected on the laser axis exhibits peaks at 1.0 MeV, 1.7 MeV and 3.8 MeV. These peaks are attributed to the N and He reactions induced by counterpropagating fast deuterons accelerated by the photon pressure of the heating pulses.

Towards a programme of testing and qualification for structural and plasma-facing materials in ‘fusion neutron’ environments

Materials damage by 14.1MeV neutrons from deuterium–tritium (D–T) fusion reactions can only be characterised definitively by subjecting a relevant configuration of test materials to high-intensity ‘fusion-neutron spectrum sources’, i.e. those simulating closely D–T fusion-neutron spectra. This provides major challenges to programmes to design and construct a demonstration fusion reactor prior to having a large-scale, high-intensity source of such neutrons. In this paper, we discuss the different aspects related to these ‘relevant configuration’ tests, including: • generic issues in materials qualification/validation, comparing safety requirements against those of investment protection; • lessons learned from the fission programme, enabling a reduced fusion materials testing programme; • the use and limitations of presently available possible irradiation sources to optimise a fusion neutron testing program including fission-neutron irradiation of isotopically and chemically tailored steels, ion damage by high-energy helium ions and self-ion beams, or irradiation studies with neutron sources of non-fusion spectra; and • the different potential sources of simulated fusion neutron spectra and the choice using stripping reactions from deuterium-beam ions incident on light-element targets.

Calculating fusion neutron energy spectra from arbitrary reactant distributions

The Directional Relativistic Spectrum Simulator (DRESS) code can perform Monte-Carlo calculations of reaction product spectra from arbitrary reactant distributions, using fully relativistic kinematics. The code is set up to calculate energy spectra from neutrons and alpha particles produced in the D(d, n)3He and T(d, n)4He fusion reactions, but any two-body reaction can be simulated by including the corresponding cross section. The code has been thoroughly tested. The kinematics calculations have been benchmarked against the kinematics module of the ROOT Data Analysis Framework. Calculated neutron energy spectra have been validated against tabulated fusion reactivities and against an exact analytical expression for the thermonuclear fusion neutron spectrum, with good agreement. The DRESS code will be used as the core of a detailed synthetic diagnostic framework for neutron measurements at the JET and MAST tokamaks.

Experimental validation of a Tokamak neutron spectrometer based on Bonner spheres**Supported by National Natural Science Foundation of China (11375195) and National Magnetic Confinement Fusion Science Program of China (2013GB104003)

In order to realize real-time fusion neutron spectrum diagnosis for the HL_2A Tokamak, a Bonner Sphere Spectrometer (BSS) array has been developed, consisting of eight polyethylene spheres (PS) with embedded 3He proportional counters. To validate its spectrometric capability, spectrum measurement of an 241Am-Be neutron source was carried out and is described. The Monte Carlo code Geant4 was used to calculate the response functions, taking this interference into consideration. Finally, the neutron spectrum was unfolded in the energy range from 10−9 MeV to 20 MeV. The unfolded spectrum has remarkable consistency with the ISO 8529-1 standard 241Am-Be neutron spectrum which is a preliminary demonstration that this BSS is reliable and practical.

Spectra of neutrons from a beam-driven fusion source

Semianalytical calculations of fusion neutron spectra in case of the presence of substantial suprathermal tails in fuel nuclei velocity distributions have been implemented using two different approaches. Numerical techniques are described. Results for a variety of isotropic and anisotropic distributions of suprathermal fuel nuclei are presented. Verification of calculated fusion product spectra has been performed using several methods.

Analysis of primary damage in silicon carbide under fusion and fission neutron spectra

Irradiation parameters on primary damage states of SiC are evaluated and compared for the first wall of ITER under deuterium–deuterium (DD) and deuterium–tritium (DT) operation, the high temperature gas-cooled reactor (HTGR) and high flux isotope reactor (HFIR). With the same neutron fluence, the studied fusion spectra produce more damage and much higher gas production than the fission spectra. Due to comparable gas production and similar weighted primary recoil spectra, HFIR is considered suitable to simulate the neutron irradiation in an HTGR. In contrast to the significant differences between the weighted primary recoil spectra of the fission and the fusion spectra, the weighted secondary recoil spectra of HFIR and HTGR match those of DD and DT, indicating that displacement cascades by the fission and the fusion irradiation are similar when the damage distribution among damaged regions by secondary recoils is taken into account.

Materials R&D for a timely DEMO: Key findings and recommendations of the EU Roadmap Materials Assessment Group

The findings of the EU Fusion Programme's ‘Materials Assessment Group’ (MAG), assessing readiness of Structural, Plasma Facing (PF) and High Heat Flux (HHF) materials for DEMO, are discussed. These are incorporated into the EU Fusion Power Roadmap [1], with a decision to construct DEMO in the early 2030s.The methodology uses project-based and systems-engineering approaches, the concept of Technology Readiness Levels, and considers lessons learned from Fission reactor material development. ‘Baseline’ materials are identified for each DEMO role, and the DEMO mission risks analysed from the known limitations, or unknown properties, associated with each baseline material. R&D programmes to address these risks are developed. The DEMO assessed has a phase I with a ‘starter blanket’: the blanket must withstand ≥2MWyrm−2 fusion neutron flux (equivalent to ∼20dpa front-wall steel damage). The baseline materials all have significant associated risks, so development of ‘Risk Mitigation Materials’ (RMM) is recommended. The R&D programme has parallel development of the baseline and RMM, up to ‘down-selection’ points to align with decisions on the DEMO blanket and divertor engineering definition. ITER licensing experience is used to refine the issues for materials nuclear testing, and arguments are developed to optimise scope of materials tests with fusion neutron (‘14MeV’) spectra before DEMO design finalisation. Some 14MeV testing is still essential, and the Roadmap requires deployment of a ≥30dpa (steels) testing capability by 2026. Programme optimisation by the pre-testing with fission neutrons on isotopically- or chemically-doped steels and with ion-beams is discussed along with the minimum 14MeV testing programme, and the key role which fundamental and mission-oriented modelling can play in orienting the research.

Diagnosing suprathermal ion populations in Z-pinch plasmas using fusion neutron spectra

The existence of suprathermal ion populations gives rise to significant broadening of and modifications to the fusion neutron spectrum. We show that when this population takes the form of a power-law at high energies, specific changes occur to the spectrum which are diagnosable. In particular, the usual Gaussian spectral shape produced by a thermal plasma is replaced by a Lorentz-like spectrum with broad wings extending far from the spectral peak. Additionally, it is found that the full width at half maximum of the spectrum depends on both the ion temperature and the power-law exponent. This causes the use of the spectral width for determination of the ion temperature to be unreliable. We show that these changes are distinguishable from other broadening mechanisms, such as temporal and motional broadening, and that detailed fitting of the spectral shape is a promising method for extracting information about the state of the ions.

Experimental studies on pulsed neutron sensitivity of a fusion neutron spectrometer

A sensitivity optimized magnetic proton recoil (MPROS) spectrometer for measuring both steady and pulsed fusion neutron spectra in several complicated environments is developed. The proton energy-position projection relationship of the spectrometer is measured by utilizing an imaging plate. The energy response to neutron sensitivity of the MPROS spectrometer is studied through a single event count method on an accelerated steady DT neutron source and simulations by a three-dimensional charged particle transport code. The detection efficiency of the spectrometer to DT neutrons reaches a level of 2×10-5 cm2 through high efficiency parameter settings, therefore the experimental data of high statistic accuracy are obtained on a comparatively weak neutron source. Results from simulations and experiments, such as α particle calibrations, and DT neutron calibrations, achieve good consistency. Based on this conclusion, precise solution technique of measured spectra can be developed to increase the sensitivity and energy resolution of measurements of pulsed neutron spectroscopy.

High-resolution spectroscopy used to measure inertial confinement fusion neutron spectra on Omega (invited)

The areal density (ρR) of cryogenic DT implosions on Omega is inferred by measuring the spectrum of neutrons that elastically scatter off the dense deuterium (D) and tritium (T) fuel. Neutron time-of-flight (nTOF) techniques are used to measure the energy spectrum with high resolution. High signal-to-background data has been recorded on cryogenic DT implosions using a well-collimated 13.4-m line of sight and an nTOF detector with an advanced liquid scintillator compound. An innovative method to analyze the elastically scattered neutron spectra was developed using well-known cross sections of the DT nuclear reactions. The estimated areal densities are consistent with alternative ρR measurements and 1-D simulations.