Measurement of (n, n′) reaction cross-sections of 79Br, 90Zr, 197Au and 207Pb with pulsed d-D neutrons

Activation cross-sections for the (n, n′) reaction were measured by means of the activation method at neutron energies of 3.1 and 2.54 MeV using a pulsed neutron beam. The target nuclei were ⁷⁹Br, ⁹⁰Zr, ¹⁹⁷Au, and ²⁰⁷Pb whose half-lives were between 0.8 and 8 s. The value of the ⁹⁰Zr(n, n′) ^90mZr reaction was obtained for the first time. In order to confirm the pulsed neutron beam measuring method, the cross-section data of ⁷⁹Br and ¹⁹⁷Au were compared with previous data obtained using a pneumatic sample transport system. The results of this comparison were in agreement within the range of experimental error. The d-D neutrons were generated by bombarding a deuterated titanium target with a 350-keV d⁺-beam at the 80° beam line of the Fusion Neutronics Source (FNS) at the Japan Atomic Energy Research Institute. In order to obtain reliable activation cross-sections, careful attention was paid to correct the efficiency for a volume source, and the self-absorption of gamma rays in irradiated samples. The systematics of the (n, n′) reaction at a neutron energy of 3.0 MeV, which can predict cross-section of (n, n′) reaction with an accuracy of 50%, was proposed for the first time on the basis of our data.     

Evaluation of neutron nuclear data for Technetium-99

Neutron nuclear data of ⁹⁹Tc was evaluated, considering cross-sections and spectra provided from recent experiments. The evaluation was made in the incident neutron energy range from 1 keV to 20 MeV, using the optical model and nuclear reaction models. The optical model calculation based on the coupled-channels method was performed for the interaction of neutrons with ⁹⁹Tc, and potential parameters appropriately chosen reasonably explain the measured data of total cross-section. The cross-section of inelastic scattering, capture, (n, 2n), (n, p), (n, α) and (n, n′α) reactions, and γ-ray emission spectra were calculated on the basis of statistical model with preequilibrium and direct components, and they were compared with available experimental data. It is found that the presently evaluated cross-sections and γ-ray emission spectra well reproduce those experimental values and that there is a large discrepancy among the present result and evaluated data for neutron emission spectra. The obtained capture cross-section increases at the energies below 1 MeV, relative to that in JENDL-4.0. This makes the transmutation efficiency of ⁹⁹Tc into stable ¹⁰⁰Ru by accelerator driven system enhanced. The production cross-section of ⁹⁹Mo important for the medical use of nuclear diagnostics reduces by 5–30% at the energies above 12 MeV, compared with JENDL-4.0.     

Overview of the US–Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. A recent MIT report (PSFC/RR-10-4, An assessment of the current data affecting tritium retention and its use to project towards T retention in ITER, Lipschultz et al., 2010) summarizes the observations from high-energy ion bombardment studies and illustrates the saturation trend in deuterium concentration due to damage from ion irradiation in tungsten and molybdenum above 1 displacement per atom (dpa). While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in plasma facing components (PFCs), it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high primary knock-on atom (PKA) energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influences the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron-irradiation. Under the framework of the US–Japan TITAN program, tungsten samples (99.99 at.% purity from A.L.M.T. Co.) were irradiated by fission neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory (ORNL), at 50 and 300 °C to 0.025, 0.3, and 2.4 dpa, and the investigation of deuterium retention in neutron-irradiated tungsten was performed in the Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe²⁺, 20 MeV W²⁺, and 700 keV H^?) with that from neutron-irradiated tungsten to identify the similarities and differences among them.     

Evaluation of neutron-induced and γ-ray-production cross sections for elemental sulfur

A new evaluation for natural S is described, this was a completely new evaluation done for ENDF/B-V. The secondary neutron cross sections and angular distributions as well as the secondary neutron energy distribution for the continuum inelastically scattered neutrons have been included. In addition, the isotopic cross sections for the various (n, particle) reactions were evaluated with the help of experimental data and model-calculated results to construct the corresponding natural S files. Precompound effects were included. The least abundant isotope ³⁶S was not considered in the evaluation. The resolved resonance region for natural S extends from 1.0 × 10^?5 eV to 1091.0 keV, and the resonance parameters for the three S isotopes ³²S, ³³S and ³⁴S have been evaluated from experimentally determined resonance parameters. The Karlsruhe total S cross-section data extending up to 20 MeV were used to construct a smooth cross-section file. Measurements due to other investigators were also used in the evaluation.     

Parameters optimization in a hybrid system with a gas dynamic trap based neutron source

A task of long-lived transuranic isotopes utilization is considered to be one of the urgent problems for the nuclear reactor technology. Using sub-critical hybrid systems is a possible solution of the problem. Budker Institute of Nuclear Physics SB RAS together with Nuclear Safety Institute RAS is working on a hybrid system with a neutron source based on the gas dynamic trap and sub-critical fuel blanket. This article presents the results obtained from a series of numerical experiments aimed at estimating the optimal system. Particularly, maximum neutron source emission rate has been estimated to reach 1 × 10¹⁸–2 × 10¹⁸ neutrons/s at the input parameters typical for such a system. Pb–Bi buffer zone impact on integral characteristics of fuel blanket has been considered. Decrease in amount of secondary fission neutrons as the result of buffer zone thickening has been revealed.The codes developed to conduct the investigations are also described in the article. The first one, GENESYS, is a zero-dimensional code aimed at modelling plasma processes in the gas dynamic trap. The second one, NMC (Neutron Monte-Carlo), is a Monte-Carlo particle transport code and is developed as a multipurpose tool for neutron transport calculation.     

Stopping cross-section and energy straggling of protons in SiC obtained by nuclear resonant reaction of protons with 12C and 28Si

In order to evaluate stopping cross-section and energy straggling of protons in compound material SiC and its constituents C and Si, resonant backscattering spectra have been measured using proton beams in an energy range 4.9–6.1 MeV per a 100 keV step. We have observed two sharp nuclear resonances at proton energies of 4.808 MeV by ¹²C and 4.879 MeV by ²⁸Si. By systematic analyses of the resonance peak profiles, i.e., energy shift of the peak position and broadening of the peak width, the values of the stopping cross-section and the energy straggling have been deduced to be compared with SRIM-2006 and Bohr’s prediction.     

Comparative study of actinide and fission product inventory of HEU and potential LEU fuels for MNSRs

A comparative study of fuel burnup and buildup of actinides and fission products for potential LEU fuels (UO₂ and U–9Mo) with existing HEU fuel (UAl₄–Al, 90% enriched) for a typical Miniature Neutron Source Reactor (MNSR) has been carried-out using the WIMSD4 computer program. For the complete burnup, the UAl₄–Al, UO₂ and U–9Mo based systems show a total consumption of 6.89, 6.83 and 6.88 g of ²³⁵U, respectively. Relative to 0.042 g ²³⁹Pu produced in case of UAl₄–Al HEU core, UO₂ and U–9Mo based cores have been found to yield 0.793 and 0.799 g, respectively, indicating much larger values of conversion ratios and correspondingly high values of fuel utilization factor. The end-of-cycle activity of the HEU core has been found 2284 Ci which agrees well with value found by Khattab where as for UO₂ based and U–9Mo based LEU cores show 1.8 and 4.8% increase with values 2326 and 2394 Ci, respectively.     

Comparison of neutron and gamma irradiation effects on KU1 fused silica monitored by electron paramagnetic resonance

Electron paramagnetic resonance (EPR) studies have been carried out on KU1 fused silica irradiated with neutrons at fluences 10²¹ and 10²² n/m², and gamma-ray doses up to 12 MGy. The effects of post-irradiation thermal annealing treatments, up to 850 °C, have also been investigated. Paramagnetic oxygen-related defects (POR and NBOHC) and E′-type defects have been identified and their concentration has been measured as a function of neutron fluence, gamma dose and post-irradiation annealing temperature. It is found that neutrons at the highest fluence generate a much higher concentration of defects (mainly E′ and POR, both at concentrations about 5 × 10¹⁸ spins/cm³) than gamma irradiations at the highest dose (mainly E′ at a concentration about 4 × 10¹⁷ spins/cm³). Moreover, for gamma-irradiated samples a lower treatment temperature (about 400 °C) is required to annihilate most of the observed defects than for neutron-irradiated ones (about 600 °C).     

Investigating permeation and transport of H isotopes in tungsten by first-principles

We have investigated permeation and transport of hydrogen (H) isotopes in tungsten (W) single crystal employing first-principles calculations in junction with Fick’ law. Permeability was approximately evaluated according to the solubility and diffusion coefficient of H. The solubility for H in bulk W from present calculation is consistent with the experimental results measured by Frauenfelder. The permeation fluxes of H isotopes are examined at the different thickness of W wall. The permeation fluxes of deuterium with the W thickness of 21 μm at the temperature of 770 K and with the W thickness of 50 μm at the temperature of 893 K were 0.68 × 10¹³ atom/m²s and 0.34 × 10¹⁴ atom/m²s, respectively. The dissociation coefficients of H isotopes are also evaluated. We believe that the present first-principles combined with Fick’ law method can be also generalized to investigate permeation and transport of H isotopes in most metals since such H isotopes behaviors in most metals are similar to those of H isotopes in W.     

Activation analysis of coolant water in ITER blanket and divertor

Coolant water in blankets and divertor cassettes will be activated by neutrons during ITER operation. ¹⁶N and ¹⁷N are determined to be the most important activation products in the coolant water in terms of their impact on ITER design and performance. In this study, the geometry of cooling channels in blanket module 4 was described precisely in the ITER neutronics model ‘Alite-4’ based on the latest CAD model converted using MCAM developed by FDS Team. The ¹⁶N and ¹⁷N concentration distribution in the blanket, divertor cassette and their primary heat transport systems were calculated by MCNP with data library FENDL2.1. The activation of cooling pipes induced ¹⁷N decay neutrons was analyzed and compared with that induced by fusion neutrons, using FISPACT-2007 with data library EAF-2007. The outlet concentration of blanket and divertor cooling systems was 1.37 × 10¹⁰ nuclide/cm³ and 1.05 × 10¹⁰ nuclide/cm³ of ¹⁶N, 8.93 × 10⁶ nuclide/cm³ and 0.33 × 10⁵ nuclide/cm³ of ¹⁷N. The decay gamma-rays from ¹⁶N in activated water could be a problem for cryogenic equipments inside the cryostat. Near the cryostat, the activation of pipes from ¹⁷N decay neutrons was much lower than that from fusion neutrons.     

Analysis and application of heavy isotopes in the environment

A growing number of AMS laboratories are pursuing applications of actinides. We discuss the basic requirements of the AMS technique of heavy (i.e., above ～150 amu) isotopes, present the setup at the Vienna Environmental Research Accelerator (VERA) which is especially well suited for the isotope ²³⁶U, and give a comparison with other AMS facilities. Special emphasis will be put on elaborating the effective detection limits for environmental samples with respect to other mass spectrometric methods.At VERA, we have carried out measurements for radiation protection and environmental monitoring (²³⁶U, ^{239,240,241,242,244}Pu), astrophysics (¹⁸²Hf, ²³⁶U, ²⁴⁴Pu, ²⁴⁷Cm), nuclear physics, and a search for long-lived super-heavy elements (Z > 100). We are pursuing the environmental distribution of ²³⁶U, as a basis for geological applications of natural ²³⁶U.     

Rhenium and molybdenum coatings for dendritic tungsten armors of plasma facing components: Concept,problems, and solutions

Fusion technology development requires materials resistant to heat, erosion and activation. Tungsten (W) could fulfill these requirements for plasma facing components of fusion power plants. A method developed by SNL to enhance the thermal and dimensional stability of W requires depositing a thin coating on the surface of dendritic W armors for IFE plasma facing components. This paper examines the waste disposal rating (WDR) associated with the activation of thin coatings of Re and Mo on the ARIES-ACT W-based divertor. Our results indicate that Re coatings must be limited to fewer than 30 μm in order to classify the Re/W-based divertor as low-level waste (LLW). Mo coatings that are larger than a few microns thick exceed the LLW disposal limit. It is recommended that, in addition to limiting the Mo thickness to 25 μm, natural Mo should be tailored to remove the Mo-94, Mo-98, and Mo-100 isotopes in order to reduce the WDR of the Mo/W-based divertor below the LLW disposal limit. If the isotopic tailoring process is inefficient or costly, SNL could focus their effort on the further development of the Re dendrite coating while Mo experiments would be regarded as a comparative study only.     

Investigation of efficient 131I production from natural uranium at Tehran research reactor

Iodine-131, which has a half-life of 8.05 days, is the one of the most widely used radionuclides in medical diagnosis and treats some diseases of thyroid gland. Optimization of ¹³¹I production in Tehran research reactor (TRR) was studied by two different methods. Primarily, standard nuclear codes such as ORIGEN, WIMS and CITATION were applied and then analytical solutions technique was followed.Calculated results and experimental works in the bench scale indicate that, by irradiation of 100 g natural Uranium (UO₂) for 100 h at 3.5 × 10¹³ (n’s/cm² s) thermal neutron flux in the TRR, one can produce about 5 Ci of ¹³¹I for medical purposes, on the other hand can produce very useful radionuclides like ⁹⁹Mo and ¹³³Xe in one batch irradiation in the unique production line.     

Influence of evaluated data of fission product poisons on criticality

The main objective of this research is to study the influence of cross-section differences of fission product poisons among various newly released evaluated cross-section libraries ENDFB-VI.8, JENDL3.2, JEF2.2, IAEA, ENDFB-VII and JEFF3.1 on criticality of an MTR type research reactor. The effect of cross-sections of poisons on the reactivity was studied with the help of WIMSD and CITATION codes. Various cross-section libraries were used in SARC (System for Analysis of Reactor Core) code for the fuel cycle analysis. It was found that the negative reactivity induced due to ¹³⁵Xe for the equilibrium core is around 36.00 mk whereas for ¹⁴⁹Sm it ranges from 6.65 to 7.06 mk. The three libraries (JENDL3.2, IAEA and ENDFB-VII) resulted in small increase in the Xenon worth as compared to the other three libraries. For Samarium, JEFF3.1 gives the highest worth whereas ENDFB-VI.8 gives the least worth among the six libraries.     

Fast neutron radiative capture cross-sections in fission product isotopes of neodymium

Fast neutron radiative capture cross-sections for Nd-148 and Nd-150 have been measured at about 10 neutron energies between 0.46 and 3.44 MeV using enriched isotopes and the activation technique. Experimental results are also compared with the theoretical one using FISPRO II computer code.     

Depth-dependent magnetic characterization of Fe films on NiO(0 0 1)

We used nuclear hyperfine spectroscopies and a ⁵⁷Fe probe layer approach to study the depth-dependent magnetic properties of ultrathin Fe films on NiO(0 0 1), a system exhibiting exchange bias. Conversion electron Mössbauer spectroscopy and nuclear resonance scattering of synchrotron radiation were employed. The samples were two Fe films with a thickness (8–10 ML) slightly above the critical thickness for the onset of ferromagnetism at room temperature, in which a 2 ML-thick probe layer, enriched in the ⁵⁷Fe Mössbauer isotope, was embedded at different depths from the Fe/NiO interface. Both techniques indicate that inside the film Fe has a metallic character, while at the interface with NiO different Fe phases are present. The main conclusion is that already a few monolayers from the interface with NiO the magnetic properties of Fe are bulk-like.     

Application of laser-induced breakdown spectroscopy for characterization of material deposits and tritium retention in fusion devices

Laser-induced breakdown spectroscopy (LIBS) is discussed as a possible method to characterize the composition, tritium retention and amount of material deposits on the first wall of fusion devices. The principle of the technique is the ablation of the co-deposited layer by a laser pulse with P (power density) ≥ 0.5 GW/cm² and the spectroscopic analysis of the light emitted by the laser induced plasma. The typical spatial extension of the laser plasma plume is in the order of 1 cm with typical plasma parameters of n_e ≈ 3 × 10²² m^?3 and T_e ≈ 1–2 eV averaged over the plasma lifetime which is below 1 μs. In this study “ITER-Like” mixed deposits with a thickness of about 2 μm and consisting of a mixture of W/Al/C and D on bulk tungsten substrates have been analyzed by LIBS to measure the composition and hydrogen isotopes content at different laser energies, ranging from about 2 J/cm² (0.3 GW/cm²) to about 17 J/cm² (2.4 GW/cm²) for 7 ns laser pulses. It is found that the laser energies above about 7 J/cm² (1 GW/cm²) are needed to achieve the full removal of the deposit layer and identify a clear interface between the deposit and the bulk tungsten substrate by applying 15–20 laser pulses while hydrogen isotopes decrease strongly after the first laser pulse. Under these conditions, the evolution of the spectral line intensities of W/Al/C/hydrogen can be used to evaluate the layer composition.     

Carrier-free production of Tc at an electron accelerator

We report on the results of studies of the ⁹⁵Tc isotope production conditions in a poorly investigated ⁹⁶Ru(γ,N)⁹⁵Tc reaction at a relatively inexpensive electron accelerator. Based on the analytical model it is demonstrated the possibility for estimating the effective and peak cross-sections and the isotope yield for a given reaction in a thick production target without preliminary determination of its excitation function. For this purpose we compared specific activities of two small samples that were activated under the same conditions using the reference reaction ⁶⁸Zn(γ,p)⁶⁷Cu and the one under investigation. The experiment on simultaneous photo-activation of natural zinc and ruthenium targets was performed followed by the investigation of their isotope composition. The specific activity of the radionuclide was measured and cross-sections for the ⁹⁶Ru(γ,N)^95g,mTc reactions were determined. The yields of desired isomers and admixtures in the natural ruthenium targets of different size were estimated. It is shown, in particular, that the operating conditions of the NSC KIPT accelerator KUT-30 can provide ^95gTc yields up to 120 mCi/h.     

Hydrogen isotope exchange in tungsten: Discussion as removal method for tritium

Hydrogen isotope exchange in re-crystallized polycrystalline tungsten was investigated at 320 and 450 K. In a first step the tungsten samples were loaded with deuterium to a fluence of 10²⁴ D/m² from a low-temperature plasma at 200 eV/D particle energy. In a second step, H was implanted at the same particle energy and similar target temperature with a mass-separated ion beam at different ion fluences ranging from 2 × 10²⁰ to 7.5 × 10²³ H/m². The analytic methods used were nuclear reaction analysis with D(³He,p)α reaction and elastic recoil detection analysis with ⁴He. In order to determine the D concentration at depths of up to 7.4 μm the ³He energy was varied from 0.5 to 4.5 MeV. It was found that already at an H fluence of 2 × 10²⁰ H/m², i.e. at 1/5000 of the initial D fluence, about 30% of the retained D was released. Depth profiling of D without and with subsequent H implantation shows strong replacement close to the surface at 320 K, but extending to all analyzable depths at 450 K especially at high fluences, leading to higher release efficiency. The reverse sequence of hydrogen isotopes allowed the analysis of the replacing isotope and showed that the release of D is balanced by the uptake of H. It also shows that hydrogen does not diffuse through a region of filled traps into a region were unfilled traps can be encounter but transport is rather a dynamic process of trapping and de-trapping even at 320 K. Initial D retention in H loaded W is an order of magnitude higher than in pristine W, indicating that every H-containing trap is a potential trap for D. In consequence, hydrogen isotope exchange is not a viable method to significantly enhance the operation time before the tritium inventory limit is reached but should be considered an option to reduce the tritium inventory in ITER before major interventions at the end of an operation period.