Fusion Neutron Source Research Articles

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.

Prospects for fusion neutron sources in the Russian nuclear power industry

Prospects for fusion neutron sources in the Russian nuclear power industry

Cathode cooling effects on the neutron production rate in the glow discharge type of fusion neutron source

Fusion reactions on the cathode surface of glow discharge deuterium–deuterium fusion neutron sources contribute significantly to the neutron production rate (NPR). While the NPR shows a linear relationship with current in the low current regime, a rise in cathode temperature in the high-current regime causes stagnation of the NPR. This tendency may be caused by high-temperature-induced desorption of deuterium on the cathode. This study aims to clarify the relationship between NPR and deuterium desorption. The present study utilized a water-cooling system to prevent deuterium desorption on the cathode. A stainless-steel 304 cathode and a diamond-like carbon (DLC)-coated cathode were tested. The cooling system kept the cathode temperature below 315 K throughout the experiment. In the case of the DLC-coated cathode, the water-cooling system improves the NPR in a high-current regime (30 mA or more in the present study). At 50 kV and 60 mA, the NPRs were 1.87 × 106 and 8.39 × 105 (n/s), with and without water cooling, respectively. Furthermore, without the cooling system, the NPR correlation with the cathode temperature indicates good agreement with the estimation model of deuterium desorption on the DLC-coated cathode. This study demonstrates that suppression of deuterium desorption in the cathode improves NPR, especially in the high-current regime.

Pulsed fast neutron yield measurements based on 79mBr activation in LaBr3(Ce)

This work aims to test the concept of an activation detector for pulsed DD fusion neutron sources, based on the production of metastable Br79m within a LaBr3(Ce) scintillator crystal via (n,n′) inelastic scattering. The pulsed neutron source employed is the NX3 Plasma Focus (PF) device operated in deuterium gas, which yields about 109 neutrons per shot. A range of D2 gas pressures, from 1 to 13 mbar are used to vary the test conditions. For the sake of comparison, a beryllium fast-neutron activation detector is used simultaneously with the LaBr3(Ce), and for each NX3 PF shot we derive neutron yield values from both Be and LaBr3(Ce) detectors, denoted YnBe and YnLaBr. The two detectors are positioned in the equatorial plane (θ=90°) of the NX3 to expose them to bursts of neutrons with energies close to 2.5 MeV, to simulate a thermonuclear DD fusion source. Overall, the shot-to-shot values of YnBe and YnLaBr obtained compare reasonably well. At each D2 gas pressures the 10-shot averaged values ⟨YnBe⟩ and ⟨YnLaBr⟩ are mostly within 10% of one another; for the worst case (10 mbar) ⟨YnLaBr⟩ is 25% higher than ⟨YnBe⟩. Overall, it is concluded that LaBr3(Ce) scintillation detectors can function as a capable and readily obtainable fast-neutron activation detector for measuring neutron yields from pulsed DD fusion sources.

Evaluation of influence of wall roughness on wall turbulence using LES simulation with MSGD model for liquid Li flow inside two-staged contraction nozzle

ABSTRACT Liquid lithium (Li) jet is planned as a beam target in fusion neutron sources (FNSs), such as IFMIF, A-FNS and IFMIF-DONES. For the safety and the efficiency of such FNSs, the high-speed Li jet need to be kept stable. For investigation of the detailed flow pattern inside the Li jet, numerical approaches using CFD simulation have been required to evaluate it because of its opacity. As one of issues about the inner flow structure, the influence of wall roughness on the vortex structure and surface variation is concerned in long-term operation of actual FNSs. In our previous simulation, in order to consider wall roughness in LES simulation, MSGD model was adopted to ANSYS Fluent using User-Defined Function, and CFD simulation in a simple parallel plate flow was conducted to confirm the performance of MSGD model. In this study, firstly, MSGD model was verified by LES simulation of Li flowing on smooth and rough plates. After that, LES simulation with MSGD model for two-staged contraction nozzle, installed into Li loop of Osaka University, was conducted. It was found that MSGD model was valid and spectrum of velocity in roughness wall cases clearly had different behavior from that in smooth wall case.

Numerical analysis of the viscosity of ZrF4–BeF2 molten salts for nuclear transmutation using fusion neutron sources

This study analyzed the influence of ZrF4 on the viscosity of a ZrF4–BeF2 binary salt towards developing BeF2-based molten salts for nuclear transmutation targets. Specifically, we performed molecular dynamics simulation using a polarization ion model and evaluated the viscosity, static structure, and applicability of the Stokes–Einstein equation for ZrF4–BeF2 compared with LiF–BeF2. Our study revealed that ZrF4–BeF2 exhibits higher viscosity than LiF–BeF2 because each ionic bond's longer lifetime extends the structural relaxation time. Furthermore, it is suggested that the ZrF and BeF relaxations contribute to the viscosity mechanism of ZrF4–BeF2. Most of the elements in long-lived fission products (LLFPs) are multivalent ions such as Zr, Pd, and Sn, which bind strongly to anions and are not expected to contribute much to reducing the viscosity of BeF2 mixed salts. In contrast, Cs, also an LLFP, is a monovalent ion, and CsF can reduce the viscosity of these salts. Mixing particular LLFPs with CsF will be one of the principles in designing BeF2-based salts for transmutation.

Preliminary design and analysis activities of the deuteron accelerator target for tritium breeding unit test

A breeding blanket (BB) is a key component that multiplies neutrons and generates tritium in the fusion demonstration reactor. Evaluation of the performance and reliability of the BB is a very important engineering task, and it is going through the ITER TBM program. The Korean DEMO program needs to complement the ITER TBM program to validate the BB, and the test facility is necessary to validate the BB in the continuous wave operation. Based on this need, the Korea Institute of Fusion Energy has initiated a pre-conceptual study to verify the long-term performance and reliability of the BB.This study deals with neutronics and engineering analyses of a solid beryllium (Be) target that generates neutrons through a deuteron accelerator for the fusion neutron source. In the neutronics analysis, the yield and energy spectrum of neutrons generated according to the incident particles and energy were analyzed. As a result, in the case of 40 MeV deuteron, the yield of neutrons with 10∼20 MeV similar to the fusion neutron was the highest in the forward direction of 1.08E+15 n/s. In addition, vanadium (V) was selected as the blistering mitigation layer (BML), and the thickness of the Be target and the BML were optimized through neutronics analysis to minimize the blistering. In the engineering analysis, a preliminary conceptual design of the target to cool the nuclear heating caused by the nuclear reaction was performed. To improve the cooling capacity of the target composed of Be-V-CuCrZr with the pressurized water coolant, the design of the target was optimized and its thermal integrity was evaluated. Next, thermal stress was evaluated by coupling the temperature distribution of the target obtained through thermal analysis with structural analysis. Through this, the preliminary design concept of the target that satisfies thermal and mechanical integrity requirements was derived.

OpenMC Interpretation of FNS SINBAD Shielding Benchmark Experiments

The Fusion Neutron Source (FNS) clean benchmark experiments on tungsten, vanadium, and beryllium assemblies from the SINBAD (Shielding Integral Benchmark Archive and Database) are analyzed to experimentally validate OpenMC (version 0.14.1-dev) fusion neutronics capabilities. The assemblies were irradiated with a 14-MeV deuterium-tritium neutron source. Neutron spectra, photon spectra, reaction rates, gamma heating rates (GHRs), and tritium production rates (TPRs) are compared to measured data in the experimental assemblies and MCNP-6.2 results. In the tungsten case, slight overestimations of the experimental data were observed in the neutron spectra, and the photon spectra agreed well with the experiments. Most of the GHRs agreed with the measured data within the range of experimental uncertainty in the tungsten and vanadium assemblies. In the vanadium assembly, the calculated neutron spectra underestimated the experiments in the low energy region while the photon spectra were well calculated when compared to experiments. The most noticeable discrepancies with experimental data in the gamma heating were observed at detector positions closest to the source. For the reaction rates, notable discrepancies with experimental data were seen at the front and rear of the assemblies. Compared to experiments, the OpenMC neutron spectra were well predicted in the beryllium assembly, whereas the calculated fission reaction rate and TPRs overestimated the experiments, an observation similar to that which has been reported by other authors. The average, overall calculation-to-experiment ratio (C/E) over nine TPR and seven GHR measurements were 1.03 ± 0.20 and 0.95 ± 0.14, respectively. In the case of verification, the OpenMC results of the benchmark calculations indicated comparable accuracy to MCNP-6.2. In general, the validation exercise showed that OpenMC can be used to analyze the fusion neutronics shielding benchmark problems.

Investigation of hydrodynamic characteristics of Li jet flowing along concave flow channel using LES for A-FNS

Liquid lithium (Li) jet is planned as a beam target in fusion neutron sources (FNSs), such as IFMIF in Japan and EU, A-FNS in Japan and IFMIF-DONES in EU. We have been studying the high-speed Li jet experimentally and numerically, and Li Loop of Osaka University (LLOU) and EVEDA Li Test Loop (ELTL) have been used. For the safety and the efficiency of such actual FNSs, it is significant to understand the detailed flow structure inside the Li jet for removing heat load due to deuteron beam. At this time, ELTL was already dismantled and thus experiments using LLOU become more important for the development of A-FNS. However, LLOU has the different configuration from A-FNS. Therefore, it is important to clarify the difference of the hydrodynamic characteristics of the Li jet between LLOU and A-FNS. In this study, two types of simulation were conducted; one was the Li flow simulation inside the nozzle of ELTL and another one was the Li jet simulation of ELTL. The former results were compared with that of LLOU and the later results were verified by comparing it to the experimental results in ELTL. Then, heat transfer inside the Li jet in beam-on-target was predicted.

Investigation of applicability of long-distance LP system for Li target diagnostics in Fusion Neutron Source

The beam target for generating a fusion neutron field in the Advanced-Fusion Neutron Source (A-FNS) is currently designed as a one-side free-surface liquid lithium (Li) jet flowing along a vertical concave wall. Characteristics of its thickness fluctuation were diagnosed in the EVEDA (Engineering Validation and Engineering Design Activity) Lithium Test Loop (ELTL), which has almost the same scale as the actual target excepting its channel width. In the diagnostics, the Laser Probe (LP) system was used. However, in an actual fusion neutron source, the diagnostics need to be operated from over 10 m away with no condenser lens at a downstream point from the final mirror for directing the incident laser to the Li surface from the perspective of Li vapor and radio activation. Whereat, the long-distance LP system for the actual Li target was conceptually designed for the IFMIF, and its precision error was also verified using a specular-reflection object and a water loop. In this previous study, the precision of measurement was evaluated as sufficiently high compared with the required limitation of surface variation (±1 mm). Then, in this study, we aim to verify the possibility of long-distance measurement of thickness variation for the free-surface Li flow. In the measurement, the incident laser head was set at a distance of about 10 m from the Li surface. And, while the previous measurement setup employed the focusing lens system with a focal length of 300 mm, some focusing lenses with longer focal length were used in this study. As a consequence, the Li surface variation was successfully detected and measured though the signal intensity of the reflected laser became low. Furthermore, it was confirmed that the impact of focal lengths on the detected signal was small.

The impact of neutral beam parameters on current drive and neutron yield in DEMO-FNS tokamak

DEMO-FNS (DEMO Fusion Neutron Source) will be a hybrid reactor designed to combine fusion and fission technologies. In a hybrid reactor the power is mainly produced by a fission blanket which is exposed to the neutron flux generated by confined fusion plasma used as a Fusion Neutron Source (FNS). The steady-state operation of the FNS is easier to achieve as compared to pure fusion reactors, allowing the FNS to be more compact. The DEMO-FNS will produce 40 MW of fusion power. The neutral beam injectors (NBI) are designed to provide a steady state plasma heating, rotation, fueling, current drive and neutron generation in DEMO-FNS. Four injector units will deliver 30 MW power in deuterium with energy Eb = 500 keV. The NBI concept and main components are inspired from ITER HNBI design implementing acceleration of negative ions with their neutralization on gas.The high density of injected power in DEMO-FNS leads to a very specific operation scenario with ahigher ratio of high-energy ions in the plasma with respect to conventional tokamaks, where the mainpart of neutron flux will result from fusion between hot-tail ions and relatively coldbackground. The current drive and neutron yield produced by the tangentially injected beam in asteady-state operation are strongly affected by magnetic configuration and plasma kineticprofiles. The fast ions deposition and performance in plasma are limited by the losses associatedwith the neutral beam shape and aiming. The main channels of direct beam losses includeshine-through losses, as well as fast ion orbital losses. The detailed evaluation of NB parametersand geometry effects is a necessary step of scenario optimization.BTR code (Beam Transmission or, in earlier versions, Beam Transmission with Reionization) is intended for injector beamlines study, and BTR detailed beam model can be also applied to perform a detailed analysis of beam losses and performance in plasma. The analysis of beam in plasma shows the ways towards higher values of beam driven current (NBCD) and neutral yield; the beam-in-plasma behaviour can be tuned for a given tokamak configuration. The NB tracing model workflow is implemented in the BTOR suite (Beam in TORoids). The results for DEMO-FNS are compared with the NBI performance in ITER tokamak, as both machines are beam-driven and implement a similar NBI concept.

Integrated Model of Neutral Beam Performance in a Fusion Neutron Source Tokamak

Integrated Model of Neutral Beam Performance in a Fusion Neutron Source Tokamak

Assessment on liquid Li fire risk under humid air condition and with heat insulators

The IFMIF-like fusion neutron sources (FNSs) such as A-FNS and IFMIF-DONES have Li fire risks and the R&D activity for the Li fire risk reduction has been conducted under the IFMIF/EVEDA project. The objectives of this study are to experimentally elucidate humidity condition in airat which no Li fire ignites and the effect of the heat insulator on the Li fire ignition. The experimental set-up was designed and fabricated in order to perform Li fire experiments safely and systematically. Around 1 g of metallic Li samples were heated with and without the heat insulators as an experimental parameter of humidity in air. In this work, no ignition was observed under the less than 0.15 vol% humidity air. The heating experiments with the heat insulators clarified that the reaction between Li and the heat insulators causing Li fire ignition didn’t occur. In conclusion, the results of the present study suggest the conditions of less than 0.15 vol% humidity air and employing the heat insulator of alkaline earth silicate wool are useful for the Li target system of FNSs.

Neutronics analysis for conceptual design of target system based on a deuteron accelerator-driven fusion neutron source

A Breeding Blanket (BB), producing tritium and heat energy converted to electricity, is a critical component for the successful Fusion Demonstration Reactor (DEMO). A dedicated neutron source is required to validate the performance of the DEMO BB in a continuous long-term fusion-like environment. Therefore, a pre-conceptual design study has been carried out for the Integrated Breeding Test Facility (IBTF) to evaluate the DEMO BB performance. The IBTF is based on a 40 MeV deuteron accelerator-driven system with a maximum current of 10 mA. A solid Beryllium (Be) target with a 20 cm width and height is used to irradiate the tritium breeding unit (TBU).This study conducted a neutronics analysis of the Be target system of the IBTF. The produced neutron energy spectrum and yield were evaluated on a beam energy of 30 to 50 MeV. The analysis results show that the forward direction neutron yield with the 40 MeV deuteron is the highest at 4.92E + 15 n/s, and the neutron ratio of 10–20 MeV is also the highest. A Blistering Mitigation Layer (BML) was added to prevent the blistering effect caused by the deuteron beam within the Be target. A comparison was done for the performance evaluation of the BML candidates vanadium, palladium, tantalum, and niobium. There were no significant differences in the forward direction neutron yield for the BML options. Vanadium was the most beneficial based on the activation analysis results. Therefore, vanadium with a 0.3 mm thickness was added behind the Be target with a 5 mm thickness. In addition, neutronics analysis was conducted for the test cell shielding designs. Concrete 3.5 m thick was required to satisfy a dose of less than 100 uSv/h during the beam operation. When evaluating the dose rate during maintenance, a dose along the cooling pipes was observed due to the activation of the pipes. The neutron flux with the cooling pipes was reduced by using B4C in the maintenance cell shielding, the dose rate was below 100 uSv/h in the test cell.Deuteron and neutron transport calculations were carried out using MCNP6 with the TENDL2019 cross-section library. Activity analysis and gamma source production were performed using FISPACT-Ⅱ with the TENDL2017 cross-section library.

Beam Transmission (BTR) Software for Efficient Neutral Beam Injector Design and Tokamak Operation

BTR code (originally—“Beam Transmission and Re-ionization”, 1995) is used for Neutral Beam Injection (NBI) design; it is also applied to the injector system of ITER. In 2008, the BTR model was extended to include the beam interaction with plasmas and direct beam losses in tokamak. For many years, BTR has been widely used for various NBI designs for efficient heating and current drive in nuclear fusion devices for plasma scenario control and diagnostics. BTR analysis is especially important for ‘beam-driven’ fusion devices, such as fusion neutron source (FNS) tokamaks, since their operation depends on a high NBI input in non-inductive current drive and fusion yield. BTR calculates detailed power deposition maps and particle losses with an account of ionized beam fractions and background electromagnetic fields; these results are used for the overall NBI performance analysis. BTR code is open for public usage; it is fully interactive and supplied with an intuitive graphical user interface (GUI). The input configuration is flexibly adapted to any specific NBI geometry. High running speed and full control over the running options allow the user to perform multiple parametric runs on the fly. The paper describes the detailed physics of BTR, numerical methods, graphical user interface, and examples of BTR application. The code is still in evolution; basic support is available to all BTR users.

Fast Beam Driven Neutron Yield in Thermonuclear Neutron Source Plasmas

The thermonuclear fusion between fast (super-thermal) particles injected in plasma as a neutral beam and the ions of the background plasma is expected to be the main source of fusion neutrons in FNS (fusion neutron source) design based on tokamak. Neutral beam contribution in fusion reactivity and in the total neutron yield depends on the high-energy ion fraction in the integral energy distribution. NESTOR code [1] calculates nuclear fusion rates in the FNS plasma volume, taking into account an external source of high-energy fast ions. Neutral beam model reproduces in detail the actual beam structure in phase space at the injection port plane; while the fast ion distributions in magnetically confined plasma are calculated using a combination of slowing-down classical formulae and magnetic field topology in the tokamak chamber. Here we discuss the issues relevant to the overall neutron production and the contribution of fast ions to the neutron output in plasma.

Development of a High Sampling Rate Data Acquisition System Working in a High Pulse Count Rate Region for Radiation Diagnostics in Nuclear Fusion Plasma Research

In this study, a high sampling rate data acquisition system with the ability to provide timestamp, pulse shape information, and waveform simultaneously under a sub megahertz pulse counting rate was developed for radiation diagnostics for magnetic confinement nuclear fusion plasma research. The testing of the data acquisition system under the high pulse counting rate condition using real signals was performed in an accelerator-based deuterium-deuterium fusion neutron source (Fast Neutron Source) at the Japan Atomic Energy Agency. We found that the pulse counts acquired by the system linearly increased up to 6 × 105 cps, and the count loss at 106 cps was estimated to be ~10%. The data acquisition system was applied to deuterium-deuterium neutron profile diagnostics in the deuterium gas operation of a helical-type magnetic confinement plasma device, called the Large Helical Device, to observe the radial profile of neutron emissivity for the first time in a three-dimensional magnetic confinement fusion device. Time-resolved measurements of the deuterium-deuterium fusion emission profile were performed. The experimentally observed radial neutron emission profile was consistent with numerical predictions based on the orbit-following models using experimental data. The data acquisition system was shown to have the desired performance.

Gas composition change during operation of a compact discharge fusion neutron source with a closed deuterium supply system

A compact cylindrical fusion neutron source has been developed for various applications, such as a D–T fusion neutron source for blanket neutron transport experiments for blanket development. Currently, the operation of a discharged fusion device is mostly deuterium–deuterium (D–D) fusion, in which D2 fuel is continuously fed to the vacuum chamber and pumped out to keep its pressure constant. Deuterium–tritium (D–T) fusion operation requires a closed supply system for radioactive tritium. So, the fuel supply system has been developed using ZrCo. In this system, however, fuel D2 gas is diluted due to light hydrogen released from the inner wall of the vacuum chamber and the surface of the electrodes, which results in a decrease of neutron production rate (NPR). Therefore, the external gas supply mode was performed to remove light hydrogen from the electrode surface before the operation of the closing system. The results of the gas analysis show that H2 was not increased and the D2 ratio was maintained. As a result, the neutron production rate achieved (NPR) > 105 n/s, and the dependency of the NPR on the voltage was similar to that of the external supply mode. These results indicate that removing light hydrogen from the electrode surface is effective to suppress fuel gas dilution.

Beamline and vacuum system design of HEBT for the A-FNS accelerator

The Advanced Fusion Neutron Source (A-FNS) project is progressing at the QST Rokkasho institute, Japan. The A-FNS will conduct fusion neutron irradiation tests of materials for a fusion demonstration reactor (DEMO reactor). For these tests a high intensity deuteron beam accelerator is used to produce a high neutron flux with the neutron energies around 14 MeV. The accelerator system accelerates a deuteron beam to a beam energy of 40 MeV, guides the beam towards a liquid lithium target and shapes the uniform beam at the Li target using a magnet system.Two of the main considerations for the A-FNS high energy beam transport (HEBT) design are protection of the Superconducting Radio-Frequency linear accelerator (SRF) and vacuum system design for the interface of the HEBT and the target system. For the A-FNS, back streaming of Li vapor from the windowless free surface of the Li target is a concern. In addition, widely different vacuum pressure must be maintained between the accelerator system and the target system. This paper shows the detailed design of the A-FNS HEBT section regarding these considerations.

Influence of the Hydrogen Isotope Affinity of the Cathode Coating Material on the Neutron Production Rate in the Glow Discharge–Type Fusion Neutron Source

The glow discharge–type fusion neutron source is a compact system that generates neutrons by inducing a nuclear fusion reaction between ionized-trapped deuterium and/or tritium in the system potential well. This study aims to clarify the relationship between the neutron production rate (NPR) and the deuterium depth distribution on the cathode surface. Four units of nontransparent cathodes fabricated from stainless steel as the electrode’s base material was investigated. Two units were coated with diamond-like carbon (DLC) and titanium, which have different affinities for hydrogen isotopes, and two were uncoated units. The NPR and cathode depth profiles were determined and scanned at different operating conditions for the coated cathodes and then compared to the uncoated ones. The results revealed that the DLC-coated cathode showed much higher NPR than the other units. The increase in NPR for the system implementing a DLC-coated cathode relative to the uncoated cathode ranged from 4.7 to 10 times. In addition, the depth profile for the nontransparent cathodes showed that the deuterium concentration on/in the DLC-coated surface was more significant by about one order of magnitude than that of the other cathodes. The increase in the NPR can be attributed to the high affinity of the DLC to capture deuterium on a cathode surface. The study suggests that DLC is a promising coating for the electrode in the neutron source at low operating conditions of less than 2 kW. In the meantime, further experimental studies are planned to find more candidate materials with better performance and higher and more stable NPR as a function of time.