Deuterium-deuterium Research Articles

Potential Technosignature from Anomalously Low Deuterium/Hydrogen in Planetary Water Depleted by Nuclear Fusion Technology

Abstract Deuterium–deuterium (DD) fusion is viewed as an ideal energy source for humanity in the far future, given a vast seawater supply of D. Here, we consider long-lived, extraterrestrial, technological societies that develop DD fusion. If such a society persisted over geologic timescales, oceanic deuterium would diminish. For an ocean mass and initial deuterium/hydrogen (D/H) ratio that were Earth-like, fusion power use of only ∼10 times that projected for humankind next century would deplete the D/H ratio in ∼(a few) ×108 yr to values below that of the local interstellar medium (ISM). Ocean masses of a few percent of Earth’s would reach an anomalously low D/H in ∼106–107 yr. The timescale shortens with greater energy consumption, smaller oceans, or lower initial D/H. Here, we suggest that anomalous D/H in planetary water below local ISM values of ∼16 × 10−6 (set by Big Bang nucleosynthesis plus deuterium loss onto dust or small admixtures of deuterium-poor stellar material) may be a technosignature. Unlike SETI using radio signals, anomalous D/H would persist for eons, even if civilizations perished or relocated. We discuss the wavelengths of strong absorption features for detecting D/H anomalies in atmospheric water vapor. These are vibrational O–D stretching at 3.7 μm in transmission spectroscopy of Earth-like worlds, ∼1.5 μm (in the wings of the 1.4 μm water band) in the shorter near-infrared for direct imaging by the Habitable Worlds Observatory, and ∼7.5-8 μm (in the wings of the broad 6.3 μm bending vibration of water) for concepts like the Large Interferometer for Exoplanets.

Fusion neutron source and array of particle detectors for nondestructive interrogation of special nuclear materials

Presented herein are the outcomes of an experimental test involving a pioneering portable-active interrogation system designed for the nondestructive detection of special nuclear materials (SNMs). The system relies on the threshold energy neutron analysis concept and incorporates a portable deuterium–deuterium (DD) neutron generator producing a particle intensity of 5 × 107 n/s, coupled with three arrays of tensioned metastable fluid detectors (TMFDs) to detect secondary neutrons from the fissile material. In the presence of the fissile material, prompt fission neutrons are emitted, with an average energy of approximately 2 MeV, and around 30% of these neutrons have energies above that of the DD neutron source (2.45 MeV). The detection of a statistically significant neutron population exceeding this threshold firmly indicates the presence of SNM. TMFDs exhibit high sensitivity in efficiently detecting neutrons above the threshold while adeptly discriminating against neutrons below the threshold as well as gamma rays. This unique feature allows the interrogation system to maintain a lightweight profile without necessitating substantial shielding materials. The validation experiments involved the placement of 70 or 140 g masses of U-235 within a 1 m3 inspection volume. Measurements were carried out over 30 min intervals, repeated numerous times, both with and without U-235, at a DD neutron source intensity of 8 × 105 n/sec. Experimental count rates with natural uranium (NU) are consistently above those without NU. The probability of detection (PD) and probability of false alarm (PFA) were assessed utilizing these count rates. The DD neutron source intensity and inspection time were normalized at 5 × 107 n/sec and 90 s, respectively. The results indicated a PD of approximately 74% and 98% for detecting 70 and 140 g of U-235, respectively, with a PFA of <5%. These promising outcomes align with the specified PD (>90%) and PFA (<5%) targets outlined in ANSI standards.

Preliminary study of a compact pulsed deuterium-deuterium neutron generator with a vacuum arc ion source and a linear induction accelerator

The pulsed neutron beam combined with the time-of-flight (TOF) technique has shown great value in many scientific research and application fields. Most of the neutron-based material identification and interrogation systems require an intense pulsed neutron beam. Time resolved prompt gamma neutron activation analysis (T-PGNAA) is an effective analytical method to greatly suppress interference from the other elements or the neutron shields of the detectors. To develop T-PGNAA method in a compact accelerator neutron source, a novel intense pulsed deuterium-deuterium (D-D) neutron generator based on a linear induction accelerator (LIA) was introduced. The generator has four major components: a vacuum arc ion source with deuterated electrode, a diode accelerator structure, a self-biased voltage secondary electron suppressed structure and a deuterated metal target. The present study characterizes the performance of the neutron generator with respect to neutron yield, the ionic current, beam profile, and beam composition as a function of the acceleration voltage at various discharge currents. Monoenergetic neutrons (2.5 MeV) are produced with a yield of ∼104 n/p and a pulse width of ∼100 ns using about 200 kV of induction acceleration voltage at 10−4 Pa vacuum environment. In this study, we will discuss various physical and technical issues related to the components of the neutron generator and the operation and merits of the novel neutron generator.

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.

Optimization design for moderator, reflector, and shielding of Deuterium-Deuterium neutron source

Deuterium-Deuterium (D-D) neutron source, which could be used in both laboratory and outdoors due to its light weight, small volume and low cost, has been widely applied in thermal neutron imaging (TNI) and prompt gamma-ray neutron activation analysis (PGNAA). And the its optimization design is to obtain highly intense thermal neutrons and low dose rate around the facility with light weight and compact volume. In this paper, the genetic algorithm (GA) combined with MCNP6 was applied to globally optimize moderator, reflector, and shielding structures. As a result, the optimized thermal neutron intensity increased by 1.68 times, while the dose rate, weight and volume respectively decreased by 37%, 52%, and 27%, compared with the data based on the current design. This method could be an effective tool for the compact neutron source design.

Mechanical design of ITER radial neutron camera Ex-Port system

The Radial Neutron Camera is an ITER diagnostic designed to measure the un-collided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions, through an array of detectors covering a poloidal plasma section along collimated Lines Of Sight (LOS). It is composed by two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module of ITER Equatorial Port 1: the In-Port RNC, devoted to plasma edge coverage, and the Ex-Port RNC, devoted to the plasma core coverage. This paper presents an overview of the mechanical design of the Ex-Port RNC at the Preliminary Design Review (PDR) stage. The Ex-Port RNC is located in the Port Interspace and consists of a massive shielding structure hosting the detector units and two sets of collimators lying on different toroidal planes. The Ex-Port RNC design is presented both from the point of view of functional requirements (e.g. LOS positions and angles, radiation shielding, weight limitations) and of manufacturability. Finally, the Ex-port RNC structural integrity is assessed, and its design validated against the main loads and load combinations.

Fusion neutron diagnostics with CVD diamond detectors

The measurement of Deuterium–Deuterium (DD) and Deuterium–Tritium (DT)-fusion neutrons is crucial for plasma diagnostics at nuclear fusion facilities, such as ITER. Elevated temperatures and high radiation levels challenge radiation detectors. Chemical Vapour Deposition (CVD) diamond withstands harsh environments and is a distinguished candidate for reliable plasma instrumentation (Angelone et al., 2021 and Passeri et al., 2021). Neutron-induced nuclear reactions on Carbon nuclei of diamond substrates are the basis for fusion neutron detection with CVD diamond detectors. DD-fusion neutrons have an energy of 2.45 MeV. The corresponding response function of diamond sensors is purely elastic scattering and has a characteristic shape. DT-fusion neutrons have 14.1 MeV kinetic energy, and their response function reveals different nuclear reactions. The 12C(n, α)9Be peak is the most prominent structure in this response function. The principles of neutron diagnostics with CVD diamond detectors are discussed in this paper and the measured response functions of diamond sensors for 2.45 MeV and 14.1 MeV neutrons are presented, as well as the fusion neutron measurement using a proton-recoil telescope.

The multi-sightline compact D-D neutron spectrometers based on CLYC7-scintillator for beam ion anisotropy study in LHD

Abstract Three compact neutron emission spectrometers (CNESs) based on the novel inorganic scintillator Cs2LiYCl6:Ce with 99% 7Li-enrichment (CLYC7) were newly installed in different lines of sight in the Large Helical Device (LHD) to study the anisotropy of the velocity distribution of the beam ions. Neutron energy spectra measurements using multi-sightline CNESs were performed under neutral-beam (NB)-heated deuterium plasma. A significant Doppler shift of the deuterium-deuterium (D-D) neutron energy induced by different NB heating and different lines of sight was observed. The measured peak neutron energies were nearly consistent with those calculated by the simple two-body kinematics of the D-D reaction using the tangency radii of CNESs and NBs. These results demonstrated the capabilities of the CNESs for neutron energy spectra measurements and the study of fast-ion physics in magnetic confinement fusion research.

Design of Neutron Activation and Radiography Facilities Based on DD Generator

Design and simulation of radiation facility using deuterium-deuterium (DD) neutron generator for neutron activation analysis (NAA) and radiography have been conducted by PHITS 3.30. A cylindrical DD neutron tube (E = 2.45 MeV isotropic, 5 x 109/s) flux is surrounded by high-density polyethylene blocks which serve as the moderator. Within the moderator there are several cavities to perform neutron activation experiments. A 90 cm long beam tube is installed either radially or tangentially for radiography purposes. Monte Carlo simulations then calculate the thermal flux inside the cavities and on the end of the beam tube. The biggest thermal flux obtained in the activation chambers is about 1.95 x 109/(cm2.s) in the cavity closest to the source center. Radial beam tube delivers thermal flux of 7.86 x 103/(cm2.s), while tangential beam tube transports 1.86 x 103/(cm2.s). Although the thermal flux in the radial beam tube is higher, the fast neutron flux is also higher, about 9.60 x 103/(cm2.s). Tangential beam tube configuration can decrease fast neutron flux to only 2.00 x 102/(cm2.s). This result can serve as a preliminary study for the commisioning of radiation facilities based on compact, low-power neutron source.

Evidence of non-Maxwellian ion velocity distributions in spherical shock-driven implosions.

The ion velocity distribution functions of thermonuclear plasmas generated by spherical laser direct drive implosions are studied using deuterium-tritium (DT) and deuterium-deuterium (DD) fusion neutron energy spectrum measurements. A hydrodynamic Maxwellian plasma model accurately describes measurements made from lower temperature (<10keV), hydrodynamiclike plasmas, but is insufficient to describe measurements made from higher temperature more kineticlike plasmas. The high temperature measurements are more consistent with Vlasov-Fokker-Planck (VFP) simulation results which predict the presence of a bimodal plasma ion velocity distribution near peak neutron production. These measurements provide direct experimental evidence of non-Maxwellian ion velocity distributions in spherical shock driven implosions and provide useful data for benchmarking kinetic VFP simulations.

Performance scaling with an applied magnetic field in indirect-drive inertial confinement fusion implosions

Magnetizing a cryogenic deuterium–tritium (DT)-layered inertial confinement fusion (ICF) implosion can improve performance by reducing thermal conduction and improving DT-alpha confinement in the hot spot. A room-temperature, magnetized indirect-drive ICF platform at the National Ignition Facility has been developed, using a high-Z, high-resistivity AuTa4 alloy as the hohlraum wall material. Experiments show a 2.5× increase in deuterium–deuterium (DD) neutron yield and a 0.8-keV increase in hot-spot temperature with the application of a 12-T B-field. For an initial 26-T B-field, we observed a 2.9× yield increase and a 1.1-keV temperature increase, with the inferred burn-averaged B-field in the compressed hot spot estimated to be 7.1 ± 1.8 kT using measured primary DD-n and secondary DT-n neutron yields.

Compton Suppression System performance evaluation for measurement of VERDI activation detectors

The VERDI detector was developed for accurate neutron measurements in the plasma-facing modules of a tokamak. It comprises a low activation capsule, capable to withstand the harsh conditions of the fusion environment, containing a defined mass of metallic foils. The neutron fluence and energy spectrum are derived by analysis of the gamma lines produced by neutron activation of the metallic elements. In this work, the use of a Compton Suppression System (CSS) is investigated aiming to enhance the sensitivity of VERDI detector analysis. The CSS consists of a 40% HPGe primary detector coupled to a set of NaI secondary detectors. The apparatus was set to discard signals simultaneously recorded on both primary and secondary detectors, thus lowering the Compton continuum. The VERDI detectors were irradiated at the Joint European Torus (JET) during the 2019 Deuterium-Deuterium (DD) campaign. The CSS performance was studied by calculating peak and continuum suppression factors. The advantage introduced by Compton suppressed gamma spectrometry for each nuclide of interest is explored and the suitability of this system for VERDI detector measurements is discussed.

Effects of Neutron and Gamma Rays Combined Irradiation on the Transcriptional Profile of Human Peripheral Blood.

We studied the effects of neutrons, neutrons and γ rays, and γ rays exposures on the transcription spectrum in human peripheral blood of three healthy adult men. Samples were irradiated with 1.42 Gy 2.5-MeV neutrons, 0.71 Gy neutrons and 0.71 Gy 137Cs γ rays, and 1.42 Gy 137Cs γ rays. Transcriptome sequencing identified 56 differentially co-expressed genes and enriched 26 KEGG pathways. There are 97, 45 and 30 differentially expressed genes in neutron, neutron and γ ray combined treatment, and γ rays, respectively, and 21, 3 and 8 KEGG pathways with significant differences are enriched. Fluorescence quantitative polymerase chain reaction (qPCR) verified differential co-expression of AEN, BAX, DDB2, FDXR, and MDM2. Additionally, irradiation of AHH-1 human lymphocytes with a 252Cf neutron source at 0, 0.14, 0.35, and 0.71 Gy, fluorescence qPCR revealed a dose-response relationship for BAX, DDB2, and FDXR at dose ranges of 0-0.71 Gy, with R2 of 0.803, 0.999, and 0.999, respectively. Thus, neutrons can induce more differentially expressed genes and enrich more pathways. Combined treatment of neutrons and γ-rays may incorporate damage of both high and low LET, the genes activated by neutrons and γ rays combined are almost the combination of genes activated by neutron and γ rays combined treatment. BAX, DDB2 and FDXR are differentially expressed after irradiation by Deuterium-Deuterium (D-D) neutron source and 252Cf neutron source, so they are expected to be molecular targets of neutron damage.

Application of VERDI detectors for ITER materials activation product inventory characterization

The VERDI detector is a passive neutron detector, based on the multi-foil activation technique, developed for measurements in harsh environments, such as those encountered in fusion devices. During the 2019 Joint European Torus (JET) Deuterium-Deuterium (DD) campaign, irradiation studies of dosimetry foils and ITER material samples were performed at the JET Long Term Irradiation Station (LTIS). VERDI detectors were used to determine the neutron fluence and energy spectrum at the irradiation position of the samples. In the present work, the VERDI detector results are used as input to FISPACT-II to predict the radionuclide product inventory of the irradiated samples. The results of the calculations are compared against activities determined by gamma spectrometry. The application of the VERDI detector is demonstrated and data for the validation of activation predictions in ITER materials irradiated in a DD fusion environment are provided.

Nuclear design of a shielded cabinet for electronics: The ITER radial neutron camera case study

The Radial Neutron Camera (RNC) is a diagnostic system located in ITER Equatorial Port #1 providing several spatial and time-resolved parameters for the fusion power estimation, plasma control and physics studies. The RNC measures the uncollided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions through an array of neutron flux detectors located in collimated Lines of Sight. Signals from RNC detectors (fission chambers, single Crystal Diamonds and scintillators) need preamplification because of their low amplitude. These preamplifiers have to be as close as possible to the detectors in order to minimize signal degradation and must be protected against fast and thermal neutrons, gamma radiation and electromagnetic fields. The solution adopted is to host the preamplifiers in a shielded cabinet located in a dedicated area of the Port Cell, behind the Bioshield Plug. The overall design of the cabinet must ensure the necessary magnetic, thermal and nuclear shielding and, at the same, satisfy weight and allocated volume constraints and maintain its structural integrity. The present paper describes the nuclear design of the shielded cabinet, performed by means of 3D particle transport calculations (MCNP), taking into account the radiation streaming through the Bioshield penetrations and the cross-talk effect from the neighboring Lower and Upper Ports. We present the assessment of its nuclear shielding performances and analyze the compliancy with the alert thresholds for commercial electronics in terms of neutron flux and cumulated ionizing dose.

Comprehensive Analysis of Streaming and Shutdown Dose Rate Experiments at JET with ORNL Fusion Neutronics Workflows

Current experimental fusion systems and conceptual designs of fusion pilot plants (FPPs) are growing in complexity and size. Several radiation metrics are crucial to the safe operation of fusion machines, including neutron flux streaming through openings and the shutdown dose rate (SDDR). Most current designs of advanced experimental fusion systems—and the most probable candidates for FPPs—are based on the tokamak concept, which is prone to neutron streaming through the myriad openings needed for diagnostic and support systems. SDDR is caused by decay gamma rays from radionuclides that become activated by neutrons during the operation of a fusion system that use deuterium-deuterium (DD), tritium-tritium, or deuterium-tritium plasma. Because computational tools have become essential for determining these radiation metrics, they must be validated against reliable and applicable experimental data. Experiments at the Joint European Torus (JET) provide a unique source of experimental data for validating computational tools and nuclear data used to determine SDDR and neutron fluxes in streaming-dominated geometries. This paper presents the comprehensive analysis of the high-performance DD JET SDDR, and streaming experiments performed using Oak Ridge National Laboratory (ORNL) fusion workflows. The computational results were compared with experimental results that consist of online SDDR measurements with ionization chambers and neutron fluence streaming measurements using thermoluminescent detectors. The ratio of calculated-to-experimental SDDR values ranges from 0.6 to 2.5, and the streaming results range from 0.5 to 8.0. Future work will include analyzing the JET 2021 DTE2 campaign alongside the integration of the Shift Monte Carlo transport code into all ORNL fusion neutronics workflows.

Progress in commissioning a neutron/X-ray radiography and tomography systems at IAEA NSIL

The Nuclear Science and Instrumentation Laboratory (NSIL) is currently establishing a Neutron Science Facility (NSF) based on two compact neutron generators: Deuterium-Deuterium (DD), resulting in 2.45 MeV neutrons, and Deuterium-Tritium (DT), resulting in 14.1 MeV neutrons, with maximum source intensities up to 5 × 106 n/s and 4 × 108 n/s over 4π, respectively. Neutron/X-ray radiography and tomography are two of the applications the NSF will be equipped with. In this paper, we report the state of the radiography and tomography system, and the neutron/X-ray radiography and tomography experiments we performed. Good radiographs and tomographs are obtained with X-ray. The spatial resolution of the X-ray radiography system is measured to be about 0.1 mm. The thermal neutron radiographs and fast neutron radiographs have proved that the DD neutron generator will be beneficial to demonstrate neutron radiography capabilities for educational purposes. Additionally, the detail of the collimator optimization of the DD neutron radiography was presented.

Nuclear Analyses for the Assessment of the Loads on the ITER Radial Neutron Camera In-Port System and Evaluation of Its Measurement Performances

The radial neutron camera (RNC) is a key ITER diagnostic system designed to measure the uncollided 14- and 2.5-MeV neutrons from deuterium–tritium (DT) and deuterium–deuterium (DD) fusion reactions, through an array of detectors covering a full poloidal plasma section along collimated lines of sight (LoS). Its main objective is the assessment of the neutron emissivity/ <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\alpha $ </tex-math></inline-formula> source profile and the total neutron source strength, providing spatially resolved measurements of several parameters needed for fusion power estimation, plasma control, and plasma physics studies. The present RNC layout is composed of two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module (DSM) of ITER Equatorial Port 1 (EP01): the ex-port subsystem and the in-port one. The ex-port subsystem, devoted to the plasma core coverage, extends from the Port Interspace to the Bioshield Plug: it consists of a massive shielding unit hosting two sets of collimators lying on different toroidal planes, leading to a total of 16 interleaved LoS. The in-port system consists of a cassette, integrated inside the port plug DSM, containing two detectors per each of the six LoS looking at the plasma edges. The in-port system must guarantee the required measurement performances in critical operating conditions in terms of high radiation levels, given its proximity to the plasma neutron source. This article presents an updated neutronic analysis based on the latest design of the in-port system and port plug. It has been performed by means of the Monte Carlo MCNP code and provides nuclear loads on the in-port RNC during normal operating conditions (NOC) and inputs for the measurement performance analysis.

A new neutron time-of-flight detector for yield and ion-temperature measurements at the OMEGA Laser Facility.

A new neutron time-of-flight (nTOF) detector for deuterium-deuterium (DD)-fusion yield and ion-temperature measurements was designed, installed, and calibrated for the OMEGA Laser Facility. This detector provides an additional line of sight for DD neutron yield and ion-temperature measurements for yields exceeding 1 × 1010 with higher precision than existing detectors. The nTOF detector consists of a 90-mm-diam, 20-mm-thick BC-422 scintillator and a gated Photek photomultiplier tube (PMT240). The PMT collects scintillating light through the 20-mm side of the scintillator without the use of a light guide. There is no lead shielding from hard x rays in order to allow the x-ray instrument response function of the detector to be measured easily. Instead, hard x-ray signals generated in implosion experiments are gated out by the PMT. The design provides a place for glass neutral-density filters between the scintillator and the PMT to avoid PMT saturation at high yields. The nTOF detector is installed in the OMEGA Target Bay along the P8A sub-port line of sight at a distance of 5.3m from the target chamber center. In addition to DD measurements, the same detector can be used to measure the neutron yield and ion temperature from deuterium-tritium (DT) implosion targets in the 5 × 1010 to 2 × 1012 yield range. The design details and the calibration results of this nTOF detector for both D2 and DT implosions on OMEGA will be presented.

DEUTERON INDUCED FUSION REACTION TARGET FOR INERTIAL CONFINEMENT FUSION (ICF)

Objective: Energy efficiency enhancement is one of the most effective ways to achieve Fast ignition (FI) in inertial confinement fusion ICF. High energy output gain is essential for ICF reactors and greater energy efficiency can reduce energy costs. The injection of Ion beam is one method used to achieve FI fusion reaction in ICF. A fusion of deuteron with lithium-6 isotope, DLi6 is reviewed in this work alongside the fusion of Deuterium – Tritium (DT), Deuterium – Deuterium (DD), Deuterium – Helium-3 (DHe3) and Proton – Boron-11 (PB11).&#x0D; Materials and Methods: In this work, it is proposed the projection of laser-driven deuteron beam in the FI scheme for ICF in the DLi6 plasma. Fusion occurs as the projected deuteron ion beam hits the lithium-6 target in the thermonuclear fusion reaction.&#x0D; Results: The results show that the fusion reactions of DD, DHe3 and PB11 all require high input kinetic energy (Mega-electronvolts) for the fusion process to occur because of higher Coulomb barrier and the probability of fusion increases by increasing the input energy drive with low output energy gain. DT fusion which require low input kinetic energy of about 400 KeV with high cross section and generated considerable high output energy gain of about 17.59 MeV, However this fusion reaction require large tritium inventory and tritium does not occur naturally, therefore the need for tritium breeding. When the energy of deuteron beam is projected at 200 keV to lithium-6 isotope target, although D + Li6 has a low total cross section of about 19.409 mbarn, the stopping power of the electrons would be more than ions, nuclear stopping power is considerable at very low deuterons energies, the Coulomb interaction of deuteron and lithium-6 occurs with output energy gain of about 22.373 MeV.&#x0D; Conclusion: The investigations indicate that fusion target energy gain efficiency is independent of lithium-6 numerical density. The highest value of energy efficiency gain occurs with lower input kinetic energy of deuteron beam of about 200 KeV to lithium-target.&#x0D; Recommendation: This findings contribute to the core mission of NIF in achieving fast ignition with low ignition energy input to achieve Lawson break-even or "ignition" point of the fusion fuel pellet, where it gives off 100% or more energy than it absorbs. However the simulation results were based on programmed model of Geant4 Hadr03. This results can be validated with the appropriate experimental design of the Hadr03 process.