MeV Neutrons Research Articles

Shielding Capability Research on Composite Base Materials in Hybrid Neutron-Gamma Mixed Radiation Fields

The 16N monitoring system operates in a mixed neutron-gamma radiation field and is subject to high background radiation, thus triggering instability in the 16N monitoring system measurement data. Due to its property of actual physical process simulation, the Monte Carlo method was adopted to establish the model of the 16N monitoring system and design a structure-functionally integrated shield to realize neutron-gamma mixed radiation shielding. First, the optimal shielding layer with a thickness of 4 cm was determined in this working environment, which had a significant shielding effect on the background radiation and improved the measurement of the characteristic energy spectrum and the shielding effect on neutrons was better than gamma shielding with the increase in the shield thickness. Then, functional fillers such as B, Gd, W, and Pb were added to the matrix to compare the shielding rates of three matrix materials of polyethylene, epoxy resin, and 6061 aluminum alloy at 1 MeV neutron and gamma energy. The shielding performance of epoxy resin as the matrix material was better than that of the aluminum alloy and polyethylene, and the shielding rate of boron-containing epoxy resin was 44.8%. The γ-ray mass attenuation coefficients of lead and tungsten in the three matrix materials were simulated to determine the best material for the gamma shielding performance. Finally, the optimal materials for neutron shielding and gamma shielding were combined, and the shielding performance of single-layer shielding and double-layer shielding in mixed radiation field was compared. The optimal shielding material-boron-containing epoxy resin was determined as the shielding layer of the 16N monitoring system to realize the integration of structure and function, which provides a theoretical basis for the selection of shielding materials in a special working environment.

63Cu(n, α)60Co cross sections in the MeV region

Cross sections of the 63Cu(n, α)60Co reaction in the MeV neutron energy region were measured using both the direct measurement method and the activation method. A twin-gridded ionization chamber was used for the direct measurement while the high-purity germanium detector was used for the activation method. The measured cross sections using the two methods are consistent within measurement uncertainties. The present results support the evaluation data in the ENDF/B-VII.1 library instead of the latest evaluation data in the ENDF/B-VIII.0 library, which is helpful to clarify discrepancies in measurement and evaluation data. TALYS-1.9 code analysis was performed which shows that the measurement cross sections can be well reproduced with minor adjustments of the input parameters. Through the calculation, it is found that the compound mechanism predominates the 63Cu(n, α)60Co reaction with ratios higher than 0.7 for neutron energy less than 20 MeV.

Neutronic simulation of medical radioisotope 99Mo and 177Lu production in IPR 14 MeV neutron generator facility

An accelerator based 14 MeV neutron generator is commissioned at Institute for Plasma Research India. The generator is based on the linear accelerator concept where the deuterium ion beam impinged to the tritium target to produce neutrons. The generator is designed to produce 5 × 1012 n/s (max.). The 14 MeV neutron source facilities are an emerging tool for the lab scale experiments and research. In order to utilize the generator for the welfare of humanity, the assessment is made for the production of medical radioisotopes using the neutron facility.The usage of radioisotopes in the treatment and diagnosis of a disease is an important factor in the healthcare sector. A series of calculations are conducted to generate radioisotopes, especially 99Mo and 177Lu those are having huge applications in the medical and pharmaceutical industries. 99Mo can be also generated through neutron reactions 98Mo(n, g)99Mo and 100Mo(n, 2n)99Mo apart from fission reaction. The cross section of 98Mo(n, g)99Mo is high in the thermal energy range whereas 100Mo(n,2n)99Mo occurs at a high energy range. 177Lu can be produced using the reactions 176Lu (n, g)177Lu and 176Yb (n, g)177Yb. The cross section of both 177Lu production routes is higher at thermal energy range. The neutron flux level near the target is around 1010 cm−2s−1. In order to enhance production capabilities, the neutron energy spectrum moderators are used to thermalize the neutrons. The materials used as a moderator are beryllium, HDPE, graphite, etc. Moderators enhance the capabilities of medical isotope production in neutron generators.

Proton induced radiation effect of SiC MOSFET under different bias

Radiation effects of silicon carbide metal–oxide–semiconductor field-effect transistors (SiC MOSFETs) induced by 20 MeV proton under drain bias (V D = 800 V, V G = 0 V), gate bias (V D = 0 V, V G = 10 V), turn-on bias (V D = 0.5 V, V G = 4 V) and static bias (V D = 0 V, V G = 0 V) are investigated. The drain current of SiC MOSFET under turn-on bias increases linearly with the increase of proton fluence during the proton irradiation. When the cumulative proton fluence reaches 2 × 1011 p⋅cm−2, the threshold voltage of SiC MOSFETs with four bias conditions shifts to the left, and the degradation of electrical characteristics of SiC MOSFETs with gate bias is the most serious. In the deep level transient spectrum test, it is found that the defect energy level of SiC MOSFET is mainly the ON2 (E c – 1.1 eV) defect center, and the defect concentration and defect capture cross section of SiC MOSFET with proton radiation under gate bias increase most. By comparing the degradation of SiC MOSFET under proton cumulative irradiation, equivalent 1 MeV neutron irradiation and gamma irradiation, and combining with the defect change of SiC MOSFET under gamma irradiation and the non-ionizing energy loss induced by equivalent 1 MeV neutron in SiC MOSFET, the degradation of SiC MOSFET induced by proton is mainly caused by ionizing radiation damage. The results of TCAD analysis show that the ionizing radiation damage of SiC MOSFET is affected by the intensity and direction of the electric field in the oxide layer and epitaxial layer.

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.

TOFu: A fully digital data acquisition system upgrade for the neutron time-of-flight spectrometer TOFOR

TOFOR is a time-of-flight (TOF) neutron spectrometer located at the Joint European Torus (JET) with a vertical sightline of the JET plasma. It consists of 5 start (denoted S1) and 32 stop (denoted S2) plastic scintillation detectors which can be used in coincidence to generate a TOF spectrum. Spectroscopic analysis of the neutron TOF spectra produced by the JET plasma is regularly performed to determine, e.g., the fuel ion ratio and the presence of fast ion species in the fusion plasma. TOFOR has been upgraded with a new digital data acquisition (DAQ) system, denoted TOFu, which consists of 10 waveform digitizers with a total of 40 channels, 37 of which are connected to the photomultiplier output of the different S1 and S2 detectors. This paper presents a technical overview of the TOFu system and describes the offline analysis capabilities of TOFu which were not available with the previous DAQ system. Two experimental JET discharges are studied and used to show that the signal-to-background ratio is improved by almost 200% for the 2.5 MeV neutron signal and almost 400% for the 14 MeV neutron signal using the new offline analysis capabilities.

Simulation of 14 MeV neutron-induced SEU in SRAM

This paper presents a full-physical methodology to simulate single event upset (SEU) induced by 14 MeV neutron in static random access memory (SRAM). Nuclear-reaction simulator TALYS is first utilized to acquire valuable information about the secondary particles generated by nuclear interactions between 14 MeV neutrons with both silicon and oxygen nuclei. Based on Geant4 code, a tool named Gproton is used exclusively for the neutron/proton-induced SEU calculation. Pioneering Gproton simulation is used to output SEU candidate events by traditional critical-charge method and computing process is accelerated through a forced collision definition. Afterwards, a commercial TCAD which has good compatibility with the input and output files of the customized Gproton code is employed to distinguish the real SEU events from the candidate SEU events randomly selected. By combining Gproton and TCAD, a calibrated cross section of 14 MeV neutron-induced SEU can be obtained. The proposed method is applied to a 180-nm commercial SRAM with six-transistor memory cell. Predicted SEU cross section is in good agreement with experimentally measured value. The results present in this paper suggest that, for SRAMs exposed in 14 MeV neutron, the combination of the material particle transporting simulation and device TCAD modeling is a candidate method to provide reasonable predictions of the cross section of SEU which improves the accuracy of the prediction effectively.

Benchmark shielding calculations for fusion and accelerator-driven sub-critical systems

We conducted benchmark calculations of neutron-shielding experiments for fusion applications such as (1) Neutron spectra in iron shields with 14 MeV neutrons; and (2) Leakage neutron spectra from various sphere piles by 14 MeV neutrons, and for Accelerator-Driven Sub-Critical system (ADS) such as (3) Neutron energy spectra after transmission through iron and concrete shields using a quasi-monochromatic neutron source, (4) Thick target neutron yield produced by high-energy heavy ion incidence on a thick target, and (5) Induced radio activity in concrete exposed to secondary particles produced by heavy-ion incident reaction on an iron target by using the Particle and Heavy Ion Transport code System (PHITS). For case (1), the calculated neutron energy spectra using the nuclear data libraries, JENDL-4.0, agreed with the experimental data in iron shields. For the cases (2) of Al, Ti, Cr, and As piles, the neutron energy spectra calculated with JENDL-4.0 agreed with the experimental results well. For the case (3), the neutron spectra calculated with JENDL-4.0/HE reproduced the experimental data because the proton data library of7Li can produce source neutrons for proton incident reactions. For the case (4), calculated results for thick target neutron spectra produced by a 400 MeV per nucleon carbon incident reaction on lead reproduced the experimental data well. For the case (5) with a neutron source produced by the 200–400 MeV per nucleon heavy-ion incident reaction, the calculated results of the reaction rate depth profiles of197Au (n, γ)198Au reactions reproduced the experimental results within a factor of 2. Overall, PHITS with nuclear data libraries reproduced the experimental data sufficiently well. Thus, PHITS has a potential application in the design of advanced reactor systems, such as fusion and ADS facilities. This experimental data are also useful in validating nuclear reaction models in other Monte Carlo codes and evaluating nuclear data libraries for advanced reactor systems.

1D/2D benchmark experiments of aluminum alloy irradiated by 14 MeV neutrons for ICF facilities

Aluminum alloy (5083) is one of the most important structural materials used in laser-driven Inertial Confinement Fusion (ICF) facilities, which will be exposed to high-yield 14 MeV neutrons and highly activated after the DT explosion. Reliable estimation of shutdown dose rate levels is of great importance for machine operation planning and to ensure the safety of people entering into target bay. Both 1D and 2D benchmark experiments were performed on a DT neutron accelerator at the Institute of Nuclear Physics and Chemistry (INPC) to validate the calculation codes and cross section libraries (JMCT with FENDL-3.1d and FISPACT with EAF-2007) used in the simulations of neutron transport and activation for aluminum alloy 5083. The total neutron yield was about 6 × 1014 in the experiments. Both radioactivity of foils placed inside the assembly and air kerma rate at given positions out of the assembly have been measured in 1D/2D experiments. Monte-Carlo code JMCT with FENDL-3.1d was used in both neutron and gamma-ray transport, and the inventory code FISPACT with EAF-2007 was used to calculate neutron activation. Results show that the relative error for air kerma rate in simulations and experiments is less than 20%, and radioisotopes give different contribution ratios on dose levels for 1D/2D experimental assemblies after DT neutron irradiation, especially for 24Na, 27Mg and 56Mn at the time within 16 h after neutron irradiated. Besides, simulations of radioactivity give good agreement with experimental results obtained for radioisotopes 24Na, 196Au, 57Ni and 58Co in both 1D and 2D experiments, and the relative error is less than ±16%.

Study of Reactor Core Loading Monitoring at the GUINEVERE Facility

The potential use of a pulsed neutron source (PNS) to measure reactivity during nuclear fuel loading as a means to prevent core loading errors has been studied at the GUINEVERE facility. This facility couples the deuteron accelerator GENEPI-3C to the fast neutron subcritical reactor VENUS-F at the Belgian Nuclear Research Center SCK·CEN. The 14-MeV neutrons are produced in the reactor core center via fusion reactions. PNS experiments were performed in five reactor configurations corresponding to the different loading steps of VENUS-F. The evolution of the neutron flux during these PNS experiments was measured by several 235U fission chambers in various positions in the inner and outer reflector and analyzed using the area-ratio method. The results show that, despite strong spatial effects, a strong correlation between the reactivity values given by the area-ratio method and some reference reactivity values remains throughout the reactor unloading. Monte Carlo simulations were first validated by comparison with the data and then used to investigate the sensitivity of the method to a core loading error. First results show that some loading errors could be experimentally detected using a PNS.

Conceptual design of a GEM (gas electron multiplier) based gas Cherenkov detector for measurement of 17MeV gamma rays from T(D, γ)5He in magnetic confinement fusion plasmas.

The only method for assessing the fusion power throughput of a deuterium-tritium (DT) reactor presently relies on determining the absolute number of 14MeV neutrons produced in the DT plasma. An independent method, developed and investigated during the recent DT campaign at the Joint European Torus, is based on the absolute counting of 17MeV gamma rays produced by the competing T(D, γ)5He reaction that features a very weak branching ratio (about 3-6 × 10-6) when compared to the main T(D,n)4He reaction. The state-of-the-art spectrometer used for gamma-ray measurements in magnetic confinement fusion plasmas is LaBr3(Ce) scintillator detectors, although they require significant neutron shielding to extract a relatively weak gamma-ray signal from a much more abundant neutron field. A better approach relies on a gamma-ray detector that is intrinsically insensitive to neutrons. We have advanced the design of a gamma-ray counter based on the Cherenkov effect for gamma-rays whose energy exceeds 11MeV, optimized to work in the neutron-rich environment of a steady-state, magnetically confined fusion plasma device. The gamma-rays interact with an aluminum window and extract electrons that move into the radiator emitting photons via the Cherenkov effect. Since the Cherenkov light consists of few photons (25 on average) in the far UV band (100-200nm), a pre-amplifier is required to transport the photons to the neutron-shielded location, which may be a few meters away, where the readout elements of the detector, either a silicon or standard photomultiplier tube, are placed. The present work focuses on the development of a scintillating GEM (Gas Electron Multiplier) based pre-amplifier that acts as a Cherenkov photon pre-amplifier and wavelength shifter. This paper presents the result of a set of Garfield++ simulations developed to find the optimal GEM working parameters. A photon gain of 100 is obtained by biasing a single GEM foil to 1kV.

Study on characteristics of neutron-induced leakage current increase for SiC power devices

In this paper, the displacement damage degradation characteristics of silicon carbide (SiC) Schottky barrier diode (SBD) and metal oxide semiconductor field effect transistor (MOSFET) are studied under 14-MeV neutron irradiation. The experimental results show that the neutron irradiation with a total fluence of 1.18×10<sup>11</sup> cm<sup>–2</sup> will not cause notable degradation of the forward <i>I-V</i> characteristics of the diode, but will lead to a significant increase in the reverse leakage current. A defect with energy level position of <inline-formula><tex-math id="Z-20230916130504">\begin{document}$ E_{\rm{C}} - 1.034 $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="18-20230976_Z-20230916130504.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="18-20230976_Z-20230916130504.png"/></alternatives></inline-formula> eV is observed after irradiation by deep level transient spectroscopy (DLTS) testing, which is corresponding to the neutron-induced defect cluster in SiC. This deep level defect may cause the Fermi level of n-type doping drift region to move toward the mid-gap level. It ultimately results in the decrease of the Schottky barrier and the increase of the reverse leakage current. In addition, neutron-induced gate leakage increase is also observed for SiC MOSFET. The gate current corresponding to <i>V</i><sub>gs</sub> = 15 V after irradiation increases nearly 3.3 times that before irradiation. The donor-type defects introduced by neutron irradiation in the oxide layer result in the difference in the conductivity mechanism of gate oxygen between before and after irradiation. The defects have an auxiliary effect on carrier crossing the gate oxide barrier, which leads to the increase of gate leakage current. The defects introduced by neutron irradiation are neutral after capturing electrons. The trapped electrons can cross a lower potential well and enter the conduction band to participate in conduction when the gate is positively biased, thus causing the gate current to increase with the electric field increasing. After electrons captured by donor-type defects enter the conduction band, positively charged defects are left from the gate oxide, leading to the negative shift of the transfer characteristics of SiC MOSFET. The results of DLTS testing indicate that the neutron irradiation can not only cause the intrinsic defect state of SiC material to change near the channel of MOSFET, but also give rise to new silicon vacancy defects. However, these defects are not the main cause of device performance degradation due to their low density relative to the intrinsic defect’s.

Cross section measurements of low-energy charged particle induced reactions using moderated neutron counter arrays

A high-and-flat efficiency moderated neutron detection array of 3He counters has been recently developed for photoneutron cross section measurements at the future ELI-NP gamma-ray beam source. We have designed a three rings geometry of 31 counters with a ˜37% efficiency flat within 5% in the 10 keV to 5 MeV neutron energy interval. A commissioning experiment was performed using proton beams delivered by the 9 MV Tandem accelerator of IFIN-HH. We measured the neutron production cross sections for proton induced reactions on nat Cu and 27Al. The (p, xn) reactions on nat Cu were investigated in the 4.5 MeV to 14 MeV proton energy range with 250 keV steps, using pulsed and continuous proton beams. The low energy measurements below 10 MeV served to validate the detection efficiency calibration against well-known natCu(p, n) cross sections. The 27Al(p, n) reaction was investigated in the 5.8 MeV to 6.75 MeV energy range with 5 keV steps. Preliminary cross section results are here compared with preceding data.

Design study of benchmark experiment for large-angle scattering cross section for non-solid target with 14 MeV neutron

Accuracy of large-angle scattering cross section in nuclear data has a large contribution on precision of neutron transport calculation in fusion reactor design. In the previous research, benchmark experiments for a solid target were carried out, however, non-solid targets, which are enclosed in a container, could not be dealt with. This is because we were not able to remove the effect due to existence of the container in the previous method. In this study, we performed design study of advanced benchmark experiment for large-angle scattering cross section especially for a non-solid target in a container. In addition, we also carried out benchmark experiments for silicon, which is important for the fusion reactor, however, is one of the elements that are difficult to obtain a solid target. In conclusion, we successfully developed an advanced benchmark experimental method for non-solid targets and verified it numerically by Monte Carlo calculation. In addition, we also found experimentally that large-angle scattering cross section of silicon is underestimated in JENDL-4, ENDF-B/VIII and JEFF-3.3.

X-Ray and Fast Neutron Radiographic Performance of an Ir-Bi-Plastic Array

Organic plastic scintillators offer the potential for exciting improvements in particle interaction efficiency, light yield (LY), and pulse shape discrimination (PSD). An area of particular application for select plastics exists in portable dual-particle imaging (DPI) systems supporting nuclear safety, security, and safeguards missions. One promising material, an Ir-Bi-Plastic, is experimentally evaluated within separate 370 kVp X-ray and 14.1 MeV neutron fields in the form of a pixelated array, which was fabricated by Lawrence Livermore National Laboratory (LLNL). Ir-Bi-Plastics pair high bismuth loadings with iridium-complex fluorophores to provide high intrinsic efficiencies for both photoelectric absorption and fast neutron detection, as well as improved LYs from the triplet-harvesting capabilities of the iridium-complex. These attributes make Ir-Bi-Plastics ideal candidates for DPI applications. Radiographic performance results for the Ir-Bi-Plastic array are compared against similar evaluations of smaller arrays made from EJ-200 and EJ-256, both industry standard materials. Measurements were recorded by pressure-coupling the three arrays to a commercial a-Si digital radiographic panel, and ASTM standard test methods and practices were utilized to calculate the modulation transfer function (MTF) and limits of spatial resolution for all three arrays. The results suggest the Ir-Bi-Plastic array, despite its larger pixel pitch, provides significantly improved performance over equivalent arrays made from either EJ-200 or EJ-256 for all X-ray and fast neutron environments practical for portable DPI.

Fusion product diagnostics based on commercially available chemical vapor deposition diamond detector in large helical device

Fusion product diagnostics based on four commercially available single-crystal chemical vapor deposition (s-CVD) diamond detectors are installed in the Large Helical Device (LHD) in order to understand energetic ion confinement. Characteristics of s-CVD diamonds were surveyed using alpha and D-T neutron sources. It is found that the energy resolutions of s-CVD diamonds for ∼ 5 MeV alpha particles and 14 MeV neutrons are 1%–3% and ∼ 1.7%, respectively. Moreover, the response of four s-CVD diamond detectors to alpha particles and D-T neutrons is almost identical. The installation positions of the diamond detectors in the vacuum vessel are searched for, based on the loss points of charged fusion products reckoned by Lorentz orbit calculations. Energy- and time-resolved measurement of fusion product flux will progress in further understanding of energetic ion confinement in LHD.

Modelling of a SiC Based Detector for the Interpretation of 14.1 MeV Neutrons Measurements

Wide-bandgap semiconductor-based fast neutron detectors such as silicon carbide (SiC) seem to be a promising concept to meet the implementation requirements as well as the performance specifications for fusion and fission environments such as radiation hardness as well as thermal and mechanical stabilities. Beyond the problem of integration of the device in constrained environments, the issue of the quantitative response of the device in a mixed radiation field must be addressed. Therefore, the characterisation of the detector performances, according to the energy of the neutrons is a key first step needed for neutron detection and monitoring in mixed radiation environments with high levels of gammas and fast and thermal neutron fluxes emissions at large energy scales. This paper presents the interpretation of 14.1 MeV neutron measurements using a numerical model developed with the GEANT4 MonteCarlo transport code. The methodology used for a first attempt to estimate the fluence rate of incident neutrons is exposed and a calculation versus experiment ratio is determined.

First characterization results of ARCADIA FD-MAPS after X-ray irradiation

The ARCADIA collaboration is developing fully-depleted (FD) Monolithic Active Pixel Sensors (MAPS) in a 110 nm CMOS process in collaboration with LFoundry. The sensor design incorporates an n+ collection node within a n-type epi-layer on top of a high-resistivity n-type substrate and p+ backside. Thus, the pn-junction sits on the backside and through an applied backside bias, the full substrate gets depleted. The targeted applications of this technology range from future high energy physics experiments to space applications, and medical and industrial scanners. Together, these applications set the minimum requirements on the detector: data collection at hit rates of (10–100) MHz/cm2, full signal processing within (1–10) μs, maximum power consumption (5–20) mW/cm2 and radiation tolerances of up to 3.4 Mrad or 6.2 × 1012 1 MeV neutron equivalent fluence. In order to proof the performance of the technology, a demonstrator chip of 512 × 512 pixels with 25 μm pitch was designed and fabricated in a first engineering run in 2021, together with additional test structures of pixel and strip arrays with different pitches and sensor geometries. The production run has produced functional passive and active pixel matrices. Earlier studies have shown that positive oxide charges and traps at the Si-SiO2 interface, introduced by ionising radiation, affect the depletion region around the collection electrode, increasing the pixel capacitance. By varying the gap size between collection node and pwells, the geometry can be optimised to keep the capacitance low also after irradiation. To study the performance after irradiation, of the optimised diode designs, the passive pixel matrices were irradiated with doses up to 10 Mrad (SiO2) using a X-ray tube with a Tungsten anode. The measurements are complemented by TCAD simulations. The maximum capacitance increase after irradiation was found to reach 6 and 12 fF/pixel for pixel pitches of 25 and 50 μm, respectively. The relative capacitance increase after irradiation has hereby been found to reach up to 250% after a dose of 10 Mrad.

Activation cross-sections for short-lived reaction products on hafnium isotopes induced by 1 – 20 MeV neutrons

Results of new activation cross-section measurements for production of 178m1Hf (T1/2 = 4.0 s) and 179m1Hf (T1/2 = 18.67 s) are presented for the following reactions: 178Hf(n,n´)178m1Hf, 179Hf(n,2n)178m1Hf, 180Hf(n,3n)178m1Hf, 179Hf(n,n´)179m1Hf, and 180Hf(n,2n)179m1Hf. The irradiations were carried out at the 7-MV Van de Graaff accelerator at EC-JRC, Geel. Neutrons in the 1-3 MeV energy range were produced via the 3H(p,n)3He reaction. Deuteron beam and a deuterium gas target were used to produce 5 and 6 MeV neutrons. For the production of quasi-monoenergetic neutrons between 16 and 19.5 MeV the 3H(d,n)4He reactions was employed. Both samples with natural composition and isotopic enrichment were employed to differentiate reactions leading to the same product. An automated pneumatic system was used for the sample irradiation, transport and radioactivity measurements. The radioactivity of the samples was determined by standard gamma-spectrometry using HPGe detector. The results obtained in the present work are compared with the data from other authors and TENDL-2017 evaluation.

Activation cross sections for reactions induced by 14 MeV neutrons on natural titanium

Cross sections for the neutrons around 14 MeV interaction with natural titanium were precisely measured by neutron activation and off-line measurement technique. The fast neutrons were produced by 3H(d,n)4He reaction and the neutron energy was obtained by using the cross section ratio method of 90Zr(n,2n)89Zr to 93Nb(n,2n)92mNb reactions. Experimental cross sections have been acquired for natTi(n,x)46Sc, natTi(n,x)47Sc, 50Ti(n,x)47Ca and 48Ti(n,x)48Sc reactions. The measured cross section data are compared with the experimental data available in the previous literature and evaluated nuclear data from the ENDF/B-VIII.0, JEFF-3.3, JENDL-5, BROND-3.1, CENDL-3.2 and FENDL-3.2b libraries. Furthermore, excitation functions for these reactions were calculated by using the theoretical model based on Talys-1.96 code with default and adjusted parameters. Within experimental error, evaluated nuclear data are mostly consistent with experimental data. The excitation function with adjusted parameters can roughly reproduce the experimental data.