MeV Fusion Neutrons Research Articles

Investigation of Cs27LiYCl6:Ce scintillator energy response for D-D fusion neutron spectrometer

During plasma experiments on EAST (Experimental Advanced Superconducting Tokamak), the discharges heated with neutral beam injection (NBI) for plasma results in the production of 2.45 MeV deuterium-deuterium (D-D) fusion neutrons. In order to measure the energy spectrum of D-D fusion neutrons, a new Cs27LiYCl6:Ce (CLYC-7) crystal with 7Li-enrichment has been developed and fabricated into a CLYC-7 scintillation detector. Before utilizing this detector for studying neutron energy spectra, it is crucial to explore the response function of the detector within a relevant neutron energy range. Therefore, the detector has been experimented using the 4.5 MeV electrostatic accelerator at Peking University to investigate its response to fast neutrons, energy resolution, and pulse discrimination capability in the neutron energy range 2.07 MeV–5.51 MeV. The neutron reactions occurring within the CLYC-7 scintillator were simulated using MCNPX, focusing on the reactions 35Cl(n,p) and 35Cl(n,α), as well as the energy response curve and the contributions of different particles (protons, alphas) to the neutron spectrum. The experimental results were successfully validated on EAST, where the 2.45 MeV fusion neutron energy spectrum was measured. That demonstrates the ability of the CLYC-7 scintillation detector to diagnose the fusion neutron energy spectrum for deuterium plasma discharge.

Effect of 14.1 MeV Fusion Neutron Irradiation on YBCO Thin Films and Commercial REBCO Tapes

Effect of 14.1 MeV Fusion Neutron Irradiation on YBCO Thin Films and Commercial <i>RE</i>BCO Tapes

Tritium removal from JET-ILW after T and D–T experimental campaigns

After the second Deuterium–Tritium Campaign (DTE2) in the JET tokamak with the ITER-Like Wall (ILW) and full tritium campaigns that preceded and followed after the DTE2, a sequence of fuel recovery methods was applied to promote tritium removal from wall components. The sequence started with several days of baking of the main chamber walls at 240 °C and at 320 °C. Subsequently, baking was superimposed with Ion-Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) cleaning cycles in deuterium. Diverted plasma operation in deuterium with different strike point configurations, including a Raised Inner Strike Point (RISP) configuration, and with different plasma heating—Ion Cyclotron Resonance Frequency (ICRF) and Neutral Beam Injection (NBI)—concluded the cleaning sequence. Tritium content in plasma and in the pumped gas was monitored throughout the experiment. The applied fuel recovery methods allowed reducing the residual tritium content in deuterium NBI-heated plasmas to about 0.1% as deduced from neutron rate measurements. This value is well below the requirement of 1% set by the maximum 14 MeV fusion neutron budget allocated in the ensuing deuterium plasma campaign. The quantified tritium removal over the course of the experiment was 13.4±0.7×1022 atoms or 0.67±0.03 g with ∼58% attributed to baking, ∼12.5% to ICWC, ∼26% to GDC, and ∼3.5% to first low power RISP plasmas. The experimentally estimated amount of removed tritium is in good agreement with long-term tritium accounting by the JET tritium reprocessing plant, in which the unaccounted amount was reduced by 0.71 g after the cleaning experiment.

Analytical and MonteCarlo approaches to infer the total gamma ray emission from the JET tokamak

A single gamma-ray spectrometer installed at the end of a collimator can beused to infer the total emission from a tokamak plasma if the transport ofgamma-rays from the plasma to the detector is known. In such analysis, theplasma emission profile plays a fundamental role, since it impacts thefraction of plasma volume intercepted by the detector line of sight.In this work, the DT 17 MeV fusion gamma-rays emission profile of theJET discharge #99608 from second 46 to 48 has been estimatedboth with the TRANSP code and reconstructed through tomographic inversionbased on the neutron camera data, assuming that fusion gamma-rayshave the same profile as the 14 MeV fusion neutrons. The gamma-raytransport has been evaluated both by MonteCarlo simulations and analyticalcalculations.By combining MonteCarlo and analytical evaluations of the gamma-ray transportin different ways with the estimated radiation emission profile, we providefour different routes to determine the total gamma-ray yield frommeasurements whose results agree within better than 10%.

Simulation study on uniaxial tensile mechanical and thermal properties of rhenium tungsten alloys

As an important potential first wall material in fusion reactor, tungsten will be irradiated by 14.1 MeV fusion neutron and then transmutated to rhenium, osmium, and tantalum during service. Due to rhenium is the main of transmutation products, which will affect the mechanical and thermal properties of the tungsten matrix. Based on this, in this work, the Young's Modulus and thermal properties of W-Re alloys with different rhenium contents were calculated by using the first principle calculations. Then, the uniaxial tensile processes in [100], [110], [111] and [211] directions were simulated by molecular dynamic simulation method. The results show that the tensile strength of W-Re alloy decreases gradually with the increase of rhenium content. With the increase of the rhenium content, the heat capacity of the material will decrease, and the thermal expansion coefficient increases.

Irradiation hardening of WK-based PFM exposed to 14 MeV protons

W-based alloys are perceived as one of the most promising plasma-facing materials (PFMs). During the service of fusion reactors, W-based PFMs will be exposed to 14 MeV fusion neutrons, causing significant displacement damage and transmutation, consequently, the irradiation-induced defects will seriously deteriorate its service performance. The irradiation effect is always a critical issue in fusion reactors and concerns the performance and reliability of W-based PFMs. In this work, WK-0.05Ti alloy was irradiated under 14MeV protons, and notable irradiation hardening was observed using microhardness and nanoindentation characterization. Then, the mechanism of W-Re precipitate formation induced by K-bubble was revealed using transmission electron microscopy. Finally, the variation of hardness values at different irradiation depths was explained based on theoretical calculations and experimental data. The present work elucidates the effect of K-bubbles on the evolution of irradiation defects in WK-based PFMs during irradiation and reveals the underlying mechanisms contributing to the irradiation hardening at different depths. Moreover, it is expected to provide additional insights into the R&D of robust plasma facing components based on WK-based materials.

FUSION BREEDING AND PURE FUSION DEVELOPMENT – PERCEPTIONS AND MISPERCEPTIONS

This paper examines fusion breeding, namely the use of 14 MeV fusion neutrons to breed 233U fuel for thermal nuclear reactors. This can be accomplished much more quickly than pure fusion. It can become main component of a power architecture that is economical, environmentally sound and has little if any proliferation risk

On the definition of the deuterium-tritium ion beam parameters for the SORGENTINA-RF fusion neutron source

The SORGENTINA-RF project aims to develop a 14 MeV fusion neutron source to produce medical radioisotopes with special focus on ^{99}Mo. The facility is based on a positive ion source with an acceleration stage to produce a deuterium (D^{+}) and tritium (T^{+}) ion beam that, impacting on a titanium-coated rotating target, allows fusion reactions to take place. Maximizing the neutron production rate is one of the main issues to be addressed in the project and the optimization of some key parameters of the ion beam is of paramount importance in this regard. In this study, a methodology is discussed to reach a definition of the beam characteristics for an effective and sustainable operation of the plant. The most convenient layout that has been found out is based on a single ion source fed by a deuterium and tritium gas mixture. Eventually, a series of considerations about the operation of the ion source and fuel cycle have been drawn.

Simulation of a micron resolution capillary liquid scintillation detector for 14 MeV fusion neutrons

Aiming to improve the spatial resolution of a neutron imaging system (NIS) for 14 MeV fusion neutrons, an ideal micron resolution capillary detector filled with a high optical index liquid scintillator was simulated. A threshold for each capillary pixel and a threshold for each cluster were applied to suppress the gamma-induced background. In addition, by using a pattern recognition algorithm and an optimized Hough transform, the accuracy of determining the neutron impinging positions and the dynamic range of this detector were enhanced. For an ideal capillary array detector, the spatial resolution is expected as one capillary size of 20μm. The dynamic range of ∼1000 is reachable while the accuracy of neutron impinging position determination keeps better than 85%. The ionization quenching, light sharing and energy resolution of the detector were applied to the simulated data to understand the capillary array detector.

Preliminary Assessment of Radiolysis for the Cooling Water System in the Rotating Target of SORGENTINA-RF

The SORGENTINA-RF project aims at developing a 14 MeV fusion neutron source featuring an emission rate in the order of 5–7 × 1013 s−1. The plant relies on a metallic water-cooled rotating target and a deuterium (50%) and tritium (50%) ion beam. Beyond the main focus of medical radioisotope production, the source may represent a multi-purpose neutron facility by implementing a series of neutron-based techniques. Among the different engineering and technological issues to be addressed, the production of incondensable gases and corrosion product into the rotating target deserves a dedicated investigation. In this study, a preliminary analysis is carried out, considering the general layout of the target and the present choice of the target material.

The 64Zn-based production route to 64Cu β ± emitter using accelerator-driven 14 MeV fusion neutrons

The development of personalized medicine is important to avoid unnecessary and expensive treatments. The theranostic approach is an established tool for specific molecular targeting, both for diagnostics and therapy. Radionuclides used for theranostic radiopharmaceuticals are typically produced in nuclear fission reactors or bio-medical cyclotrons. Here, an alternative route for producing an emerging theranostic radionuclide, namely 64Cu, is investigated by using 14 MeV fusion neutrons produced by the Frascati Neutron Generator (FNG) at the ENEA Frascati Research Centre. Samples of zinc (natural and 64Zn-enriched) were irradiated and characterized, in turn being measured at high-level metrological conditions, at the Radioactivity Laboratory of ENEA-INMRI located in Casaccia Research Centre. The results achieved in the experiments are presented and compared to those obtained by Monte Carlo simulations and previously published experimental data.

Total fusion yield measurements using deuterium–tritium gamma rays

The deuterium–tritium (DT) fusion reaction, D(T,γ)5He, has a rare (4.2 ×10−5) branching ratio that releases a ∼16 MeV gamma ray instead of the more common 14.1 MeV fusion neutron. These fusion gamma rays can be used as a complementary source of measuring the net fusion yield of experiments. Fusion gamma rays have the advantage of having negligible Doppler broadening, being isotropic and less down-scattered compared to fusion neutrons. At the National Ignition Facility, the DT fusion gamma rays are measured by the Gamma Reaction History diagnostic that thresholds to only measure >10 MeV gamma rays. After removing a ∼12% contribution from neutron capture interactions in deuterium, D(n,γ)T, and carbon, 12C(n,γ)13C, the deviation between GRH measurements and neutron-based yield measurements has been reduced to 10%. Further improvements are needed for detailed physics comparison.

Metastable acceptors in boron doped silicon: evidence of the defects contributing to carrier induced degradation

This work focuses on the recombination-enhanced reactions of boron related defects in compensated silicon and their impact on the electrical performance of silicon devices. Using deep level transient spectroscopy, we measured the defect spectra in the collector of 14 MeV fusion neutron irradiated p-n-p silicon transistors during forward current injection, as well as the corresponding degradation kinetics of both current gain and leakage current on the same device. Two hole traps at + 0.43 eV and + 0.53 eV are proved to be acceptors in the lower half of the gap with significant recombination enhanced effects. These metastabe acceptors degrade the current gain by changing the space charge distribution in the p-n-p structure and are linked with carrier-induced boron-oxygen complexes responsible for the minority-carrier lifetime degradation of silicon photovoltaics.

64Cu production by 14 MeV neutron beam

64Cu is an emerging radionuclide of great interest in personalized nuclear medicine. It is produced by a cyclotron via the reaction 64Ni(p,n)64Cu. This production method increased during the last decades, because small biomedical cyclotrons can be easily installed close to the nuclear medicine department of a hospital. As a matter of fact, 64Ni is a very expensive target material. For this reason, an alternative 64Cu production method was investigated at ENEA by using the quasi-monochromatic 14 MeV fusion neutron beam made available at the Frascati Neutron Generator (FNG) located at the ENEA – Frascati Research Center. In particular, two nuclear reactions were studied: 65Cu(n,2n)64Cu and 64Zn(n,p)64Cu. The radiochemical analysis of the activated samples was performed at the ENEA-NMLNWM laboratory located in ENEA-Casaccia Research Center. The activity measurements were carried out at the ENEA-INMRI, located in the ENEA-Casaccia Research Center, with high metrological level conditions and by assuring their traceability to the 64Cu primary activity standard here developed and maintained. A prediction of the 64Cu production by means of the high-brilliance 14 MeV neutron source named Sorgentina is also discussed.

Fusion breeding for mid-century, sustainable, carbon free power

Fusion has often been billed as the ultimate 21st century sustainable energy source. However, not only is the pace of the program glacially slow, it seems to recede further and further into the future. One way to speed up the delivery of economical fusion could be to change the objective from pure fusion, that is the use of the 14 MeV fusion neutron's kinetic energy to boil water; to fusion breeding, that is the former, but also making use of the neutrons 'potential energy' to breed ten times its energy in the form of nuclear fuel to be burned in separate reactors. The requirements of a fusion breeder are greatly relaxed from the requirements for a pure fusion reactor. For instance, ITER, the large tokamak being built by an international consortium in France, could well be the basis of an economical fusion breeder, but would have to clear many more scientific and technical hurdles before it could become the basis for a pure fusion reactor; hurdles it may or may not be able to clear. Even if it clears them, ITER is unlikely to evolve into an economical pure fusion power supply this century. A fusion breeder as could be alternate approach to speed the delivery of economical of fusion power.

Tungsten–chromium–yttrium alloys as first wall armor material: Yttrium concentration, oxygen content and transmutation elements

Tungsten (W) is a prime candidate as first wall armor material of future fusion power plants as W withstands extreme particle, heat, and radiation loads without forming long-lived radioactive waste. The release of radioactive material from the reactor to the environment should be suppressed in case of an accident such as a loss of coolant (LOCA) with simultaneous air ingress into the vacuum vessel. W oxidizes and sublimates in case of a LOCA. Therefore, oxidation resistant tungsten, chromium, yttrium (W–Cr–Y) alloys are developed to provide intrinsic safety in case of such an accident. In this paper, the optimization of the yttrium (Y) concentration is presented on bulk samples compacted by field assisted sintering technology (FAST). W with 11.4 weight (wt)% Cr and 0.6 wt% Y appears to be an optimum regarding the oxidation resistance. Further, first preparations for industrial upscaling, which may increase the impurity level, are addressed. The oxygen (O) content is varied systematically. It is shown that a good oxidation resistance requires a low O level. The exposure of the material to fusion neutrons is another issue addressed on W–Cr–Y alloys. In a non-activated environment it is shown that 1 wt% rhenium (Re) dramatically changes the oxidation kinetics: at 1273 K the mass gain of W–Cr–Y–Re follows a cubic rate law while W–Cr–Y follows a linear rate law for two days. Further, the influence of the alloying elements on the neutron transport and transmutation of W is studied by simulating the exposure of spatially heterogeneous high-resolution models of the W–Cr–Y alloys to 14 MeV fusion neutrons.

Carbon ablator areal density at fusion burn: Observations and trends at the National Ignition Facility

For inertial confinement fusion experiments, the pusher is composed of a high-density deuterium tritium cyrogenic fuel layer and an ablator, often made of carbon. In an ideal, no-mix implosion, increasing the areal density of the pusher transfers more pressure to the hot spot and increases the hot spot confinement time. There has been a lack of knowledge about the final compressed state of the ablator for implosions at the National Ignition Facility. 14 MeV fusion neutrons inelastically scattering on the remaining carbon ablator excites a nuclear metastable state that emits a prompt 4.4 MeV gamma ray. The gamma reaction history diagnostic data, when reduced by a new data analysis technique, can isolate and measure the carbon gamma rays, which are proportional to the areal density of the ablator during fusion burn. The trends over many National Ignition Facility campaigns show that the ablator areal density is weakly sensitive to the maximum shell velocity, the cold fuel radius, the ablator mass remaining, or the laser picket intensity. Controlled parameter scans reveal that, for specific campaigns, ablator compression has a strong dependence on laser coast time, high Z dopants, and the laser drive foot duration. Using a model of the compressed ablator density profile reveals that the greatest variation of the ablator areal density comes from its thickness, with highly compressed, thin layers having high areal density values. The compression and thickness of the ablator are other metrics that designers should understand to differentiate the types of capsule degradation and maximize the inertial confinement fusion performance.

14 MeV neutron streaming calculations for JET-like maze entrance using TRIPOLI-4 Monte Carlo code

For the future ITER (International Thermonuclear Experimental Reactor) operation needs, D-T fusion experiments named DTE2 will be conducted in 2020 at JET (Joint European Torus). During the DTE2 operations, 14 MeV fusion neutrons will be generated. One of the aims of the DTE2 neutronics experiments is to investigate the D-T neutron streaming along the penetrations of biological shield of the JET Torus Hall. The TRIPOLI-4 Monte-Carlo radiation transport code has been extensively used on radiation shielding analyses. The purpose of this study is to evaluate the variance reduction (VR) techniques of TRIPOLI-4 on the 14 MeV neutron streaming calculations for a JET-like maze entrance. A JET-like torus hall building of 40 m x 40 m x 4.2 m and a maze entrance of 11.8 m length in 5-section configuration of concrete were first modeled. The standard INIPOND VR method and the new developed VR method AMS (Adaptive Multilevel Splitting) were then presented and discussed. Using the single-step approach combined with AMS techniques, the TRIPOLI-4 calculations results including neutron flux and ambient dose equivalent maps, the VR performance, and the user-friendly AMS input of TRIPOLI-4 code are reported.

On the slowing-down of 14 MeV fusion neutrons: A spectrometry benchmark and perspectives on future neutron science facilities

The spectral fluence rate of the thermalized neutron field obtained by slowing down the 14 MeV fusion neutrons produced at the accelerator-driven Frascati Neutron Generator is measured by means of the Bonner Sphere Spectrometer. Neutron thermalization is achieved by means of a moderator assembly made of a copper pre-moderator and a polyethylene moderator. A Monte Carlo simulation reproducing the experimental set-up is also performed by means of the MCNP code and the results are compared to the experimental data. The benchmarked Monte Carlo is then used to predict the brilliance of a thermal moderator operating at a potential high-intensity 14 MeV neutron continuous source featuring a neutron emission rate of . The results obtained show that if a D-T neutron source featuring a continuous neutron emission rate of could be made operative, it could be effectively exploited for neutron science.

Activation Inventories after Exposure to DD/DT Neutrons in Safety Analysis of Nuclear Fusion Installations

Irradiations with 14 MeV fusion neutrons are planned at Joint European Torus (JET) in DT operations with the objective to validate the calculation of the activation of structural materials in functional materials expected in ITER and fusion plants. This study describes the activation and dose rate calculations performed for materials irradiated throughout the DT plasma operation during which the samples of real fusion materials are exposed to 14 MeV neutrons inside the JET vacuum vessel. Preparatory activities are in progress during the current DD operations with dosimetry foils to measure the local neutron fluence and spectrum at the sample irradiation position. The materials included those used in the manufacturing of the main in-vessel components, such as ITER-grade W, Be, CuCrZr, 316 L(N) and the functional materials used in diagnostics and heating systems. The neutron-induced activities and dose rates at shutdown were calculated by the FISPACT code, using the neutron fluxes and spectra that were provided by the preceding MCNP neutron transport calculations.