Oscillator strengths and transition probabilities from the Breit–Pauli R-matrix method: Ne IV

The atomic parameters–oscillator strengths, line strengths, radiative decay rates (A$A$), and lifetimes–for fine structure transitions of electric dipole (E1) type for the astrophysically abundant ion Ne IV are presented. The results include 868 fine structure levels with n≤$n\le$ 10, l≤$l\le$ 9, and 1/2≤J≤$\le J\le$ 19/2 of even and odd parities, and the corresponding 83,767 E1 transitions. The calculations were carried out using the relativistic Breit–Pauli R-matrix method in the close coupling approximation. The transitions have been identified spectroscopically using an algorithm based on quantum defect analysis and other criteria. The calculated energies agree with the 103 observed and identified energies to within 3% or better for most of the levels. Some larger differences are also noted. The A$A$-values show good to fair agreement with the very limited number of available transitions in the table compiled by NIST, but show very good agreement with the latest published multi-configuration Hartree–Fock calculations. The present transitions should be useful for diagnostics as well as for precise and complete spectral modeling in the soft X-ray to infra-red regions of astrophysical and laboratory plasmas.     

Perturbation theory for nuclear fuel depletion calculations with predictor–corrector method

The perturbation theory for nuclear fuel depletion calculations with the predictor–corrector method is derived. This theory is implemented to a reactor physics code system CBZ, and the theory itself and its implementation are numerically verified. Sensitivities of nuclide number densities after fuel depletion with respect to nuclear data calculated with this theory are compared with reference sensitivities calculated by numerical differentiation, and good agreements are obtained. Importance of accurate angle integration on product of neutron flux and generalized adjoint neutron flux is also pointed out. Sensitivities in a 3×3 multi-cell system including a gadolinium-bearing fuel pin are calculated, and it is demonstrated that the derived theory yields accurate sensitivities even if coarse depletion time step division is adopted. The present work drastically increases the applicability of the depletion perturbation theory to actual problems.     

Corrosion resistance of Mo–Fe–Ti alloy for overpack in simulating underground environment

In order to examine the application of Mo–Fe–Ti alloy for overpak, the corrosion resistance of heat-treated its alloys was investigated by electrochemical impedance spectroscopy (EIS) and transmission electron microscopy (TEM). The sample subjected to solution heat treatment (ST) had a single β phase and samples subjected to aging heat treatment at 600–700 °C had α phase precipitation in β phase. EIS results showed that the corrosion resistance of the aging heat-treated samples was lower than that of the ST sample, but much higher than that of pure Ti in 10% NaCl solution of pH 0.5 at 97 °C which simulating the crevice solution. Laser micrographs of the aging heat-treated samples indicated that α phase was caused selective dissolution in test solution. The TEM combined with EDAX (energy dispersive X-ray) analyses showed that β phase matrix composed of 2.7 wt.% Mo and 4.8 wt.% Fe, and α phase composed of 0.7 wt.% Mo and 0.1 wt.% Fe in sample aged at 600 °C. Thus, Mo-poor α phase was selectively dissolved in a test solution. In EIS, the ST sample of only β phase showed the highest resistance, and aging heat-treated samples containing α phase (0.7 wt.% Mo) showed higher values than pure Ti in the corrosion test. As Fe was involved in β phase with Mo which increased remarkably the corrosion resistance, the addition of Fe did not decrease the corrosion resistance of aging heat-treated Mo–Fe–Ti alloy in simulating underground environment.     

Metallurgical characterizations of Fe–Cr–Ni–Zr base alloys developed for geological disposal of radioactive hulls

Alloy melting route is currently being considered for radioactive hulls immobilization. Towards this, wide range of alloys, belonging to Zirconium–Iron binary and Zirconium–Stainless steel pseudo-binary systems have been prepared through vacuum arc melting route. Detail microstructural characterization and quantitative phase analyses of these alloys along with interaction study between Zirconium and Stainless steel coupons at elevated temperatures identify Zr(Fe,Cr)₂, Zr(Fe,Cr), Zr₂(Fe,Cr), Zr₃(Fe,Ni), Zr₃(Fe,Cr), Zr₃(Fe,Cr,Ni), β-Zr and α-Zr as the most commonly occurring phases within the system for Zirconium rich bulk compositions. Nano-indentation studies found Zr(Fe,Cr)₂ and Zr(Fe,Cr) as extremely hard, Zr₃(Fe,Ni) as moderately ductile and β-Zr, Zr₂(Fe,Cr) as most ductile ones among the phases present. Steam oxidation studies of the alloys, based on weight gain/loss procedure and microstructural characterization of the mixed oxide layers, suggest that each of the alloys responded to the corrosive environment differently. Fe₂O₃, NiFe₂O₄, NiO, monoclinic ZrO₂ and tetragonal ZrO₂ are found to be most common constituents of the oxide layers developed on the alloys. Integrating the microstructural, mechanical and corrosion properties, ZrFeCrNi3 (Zr: 84.00, Fe: 11.20, Cr: 3.20, Ni: 1.60, in wt.%) is identified as the acceptable base alloy for disposal of radioactive hulls.     

MCOR – Monte Carlo depletion code for reference LWR calculations

The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged.     

Model reduction methods for cylindrical structures in reactor internals considering the fluid–structure interaction

Vibrational analysis of the complex structure of reactor internals using the finite element method leads to considerable computational expense. Additionally, fluid–structure interaction (FSI) effects due to liquid coolant result in a large number of fluid elements. Here, we describe a model reduction method based on Guyan theory to solve these complex numerical problems efficiently. The master degrees of freedom selection process, which is based on the shapes of vibrational modes, is discussed. We consider the structural characteristics of the cylindrical parts of the reactor, and include FSI effects. To verify the model reduction method, several numerical examples of simple cylindrical shells are described with and without the coolant. Practical application to the internals of an advanced pressurized reactor 1400 (APR1400) is discussed with various different conditions.     

Development of thermodynamic databases in the system U–Zr–Ce–Cs–Fe–B–C–I–O–H for application to simulating phase equilibria in severe nuclear accidents

In case of severe nuclear accidents involving melt down of nuclear fuels at high temperatures, it is of considerable importance to accurately evaluate the highly-volatizing behavior of fission products (FPs) over multicomponent debris. Particularly, cesium (Cs)- and iodine (I)- bearing chemical species are regarded as notable FPs. In the present work, the authors have generated original thermodynamic databases for the system U–Zr–Ce–Cs–Fe–B–C–I–O–H featuring Cs- as well as I-bearing subsystems, which are contained in oxide, iodide, and metal (including borides and carbides) sub-databases. It has been confirmed that the phase diagrams calculated by the present set of the databases reproduce the corresponding literature data well in various kinds of subsystems of the above multicomponent system. The present set of databases has subsequently been applied to simulate phase equilibria and volatizing behavior of Cs- and I-including species, respectively, in multicomponent debris under specific temperature and atmospheric conditions corresponding to severe nuclear accidents.     

Implementation of high-fidelity neutronics and thermal–hydraulic coupling calculations in HNET

To perform nuclear reactor simulations in a more realistic manner, the coupling scheme between neutronics and thermal-hydraulics was implemented in the HNET program for both steady-state and transient conditions. For simplicity, efficiency, and robustness, the matrixfree Newton/Krylov (MFNK) method was applied to the steady-state coupling calculation. In addition, the optimal perturbation size was adopted to further improve the convergence behavior of the MFNK. For the transient coupling simulat...     

Irradiation of a thorium–plutonium rodlet: Experiment and benchmark calculations

A benchmark exercise for thorium–plutonium fuel, based on experimental data, has been carried out. A thorium–plutonium oxide fuel rodlet was irradiated in a PWR for four consecutive cycles, to a burnup of about 37 MWd/kgHM. During the irradiation, the rodlet was inserted into a guide tube of a standard MOX fuel assembly. After the irradiation, the rod was subjected to several PIE measurements, including radiochemical analysis. Element concentrations and radial distributions in the rodlet, multiplication factors and distributions within the carrier assembly of burnup and power were calculated. Four participants in the study simulated the irradiation of the MOX fuel assemblies including the thorium–plutonium rodlet using their respective code systems; MCBurn, HELIOS, CASMO-5 and ECCO/ERANOS combined with TRAIN. The results of the simulations and the measured results of the radiochemical analysis were compared and found to be in fairly good agreement when the calculated results were calibrated to give the same burnup of the thorium–plutonium rodlet as that experimentally measured. Average concentrations of several minor actinides and fission products were well reproduced by all codes, to the extent that can be expected based on known uncertainties in the experimental setup and the cross section libraries. Calculated results which could not be confirmed by experimental measurement were compared and only two significant anomalies were found, which can probably be addressed by limited modifications of the codes.     

Benchmarking of the 3-D CAD-based Discrete Ordinates code “ATTILA” for dose rate calculations against experiments and Monte Carlo calculations

Shutdown dose rate (SDDR) inside and around the diagnostics ports of ITER is performed at PPPL/UCLA using the 3-D, FEM, Discrete Ordinates code, ATTILA, along with its updated FORNAX transmutation/decay gamma library. Other ITER partners assess SDDR using codes based on the Monte Carlo (MC) approach (e.g. MCNP code) for transport calculation and the radioactivity inventory code FISPACT or other equivalent decay data libraries for dose rate assessment. To reveal the range of discrepancies in the results obtained by various analysts, an extensive experimental and calculation benchmarking effort has been undertaken to validate the capability of ATTILA for dose rate assessment. On the experimental validation front, the comparison was performed using the measured data from two SDDR experiments performed at the FNG facility, Italy. Comparison was made to the experimental data and to MC results obtained by other analysts. On the calculation validation front, the ATTILA's predictions were compared to other results at key locations inside a calculation benchmark whose configuration duplicates an upper diagnostics port plug (UPP) in ITER. Both serial and parallel version of ATTILA-7.1.0 are used in the PPPL/UCLA analysis performed with FENDL-2.1/FORNAX databases. In the FNG 1st experimental, it was shown that ATTILA's dose rates are largely over estimated (by ～30–60%) with the ANSI/ANS-6.1.1 flux-to-dose factors whereas the ICRP-74 factors give better agreement (10–20%) with the experimental data and with the MC results at all cooling times. In the 2nd experiment, there is an under estimation in SDDR calculated by both MCNP and ATTILA based on ANSI/ANS-6.1.1 for cooling times up to ～4 days after irradiation. Thereafter, an over estimation is observed (～5–10% with MCNP and ～10–15% with ATTILA). As for the calculation benchmark, the agreement is much better based on ICRP-74 1996 data. The divergence among all dose rate results at ～11 days cooling time is no more than 15% among all participants.     

Enhanced radiation tolerance of ultrafine grained Fe–Cr–Ni alloy

The evolutions of microstructure and mechanical properties of Fe–14Cr–16Ni (wt.%) alloy subjected to Helium ion irradiations were investigated. Equal channel angular pressing (ECAP) process was used to significantly reduce the average grain size from 700 μm to 400 nm. At a peak fluence level of 5.5 displacement per atom (dpa), helium bubbles, 0.5–2 nm in diameter, were observed in both coarse-grained (CG) and ultrafine grained (UFG) alloy. The density of He bubbles, dislocation loops, as well as radiation hardening were reduced in the UFG Fe–Cr–Ni alloy comparing to those in its CG counterpart. The results imply that radiation tolerance in bulk metals can be effectively enhanced by refinement of microstructures.     

Investigations of the melting behaviour of the U–Zr–Fe–O system

During a severe nuclear accident, the UO₂ fuel rods, Zircaloy cladding, guide tubes, absorber and steel structural components inside the reactor pressure vessel overheat and a series of interactions between these elements and the steam atmosphere occur. These produce more heat in addition to the decay heat and result in a liquid corium of oxidic and metallic phases depending on the exact conditions and processes. A major systems resulting from this is the U–Zr–Fe–O system. High-temperature data for this system is important in order to be able to model these interactions. The Joint Research Centre, Institute for Transuranium Elements (JRC-ITU) has been examining the melting ranges for this system over the whole FeO range by means of a specialized laser flash technique that achieves very high temperatures and avoids crucible contamination. The melted zones were examined for their structure, composition and for estimation of the liquidus and solidus temperatures. The results showed that with FeO contents of over 20mol% there was a very large melting range that would permit long liquid cooling times and extend the relocation of fuel material within the reactor pressure vessel. Based on these results, the main phase regimes expected under severe accident conditions could be identified.     

Fundamental experiments on phase stabilities of Fe–B–C ternary systems

To understanding the control blade degradation mechanism of a boiling-water reactor (BWR), a thermodynamic database for the fuel assembly materials is a useful tool. Although the iron, boron, and carbon ternary system is a dominant phase diagram, phase relation data are not sufficient for the region in which boron and carbon compositions are richer than the eutectic composition. The phase relations of three samples such as Fe_0.68B_0.06C_0.26 (at%), Fe_0.68B_0.16C_0.16 (at%), and Fe_0.76B_0.06C_0.18 (at%) were analyzed by X-ray diffraction, scanning electron microscopy, and energy–dispersive X-ray spectrometry. The results indicate that the Fe₃(B,C) phase exists only in the intermediate region at 1273 K and that the solidus temperature widely maintains at approximately 1400 K for all three samples; these results differ from the calculated data using the previous thermodynamic database. The difference might originate from the overestimation of the interaction parameter between boron and carbon in Fe₃(B,C). Proper titling was performed using the present data.     

Uncertainty quantification in (α,n) neutron source calculations for an oxide matrix

We present a methodology to propagate nuclear data covariance information in neutron source calculations from (α,n) reactions. The approach is applied to estimate the uncertainty in the neutron generation rates for uranium oxide fuel types due to uncertainties on 1) ^17,18 O(α,n) reaction cross-sections and 2) uranium and oxygen stopping power cross sections.The procedure to generate reaction cross section covariance information is based on the Bayesian fitting method implemented in the R-matrix SAMMY code. The evaluation methodology uses the Reich-Moore approximation to fit the ^17,18 O(α,n) reaction cross-sections in order to derive a set of resonance parameters and a related covariance matrix that is then used to calculate the energy-dependent cross section covariance matrix. The stopping power cross sections and related covariance information for uranium and oxygen were obtained by the fit of stopping power data in the α-energy range of 1 keV up to 12 MeV.Cross section perturbation factors based on the covariance information relative to the evaluated ^17,18 O(α,n) reaction cross sections, as well as uranium and oxygen stopping power cross sections, were used to generate a varied set of nuclear data libraries used in SOURCES4C and ORIGEN for inventory and source term calculations. The set of randomly perturbed output (α,n) source responses, provide the mean values and standard deviations of the calculated responses reflecting the uncertainties in nuclear data used in the calculations. The results and related uncertainties are compared with experiment thick target (α,n) yields for uranium oxide.     

Relativistic configuration interaction calculations for the Kα and Kβ X-ray satellites of iron

Multi-configuration Dirac-Fock and relativistic configuration interaction calculations with the Breit interaction, quantum electrodynamics, and finite nuclear mass corrections were carried out in the extended optimal level scheme for the transition wavelengths, electric dipole transition rates, and oscillator strengths of the Kα and Kβ X-rays from Fe XVII through Fe XXV. The calculated values were compared with the available data on He-like, Li-like, Be-like, and F-like iron, and were in very good agreement. These data provide reference values for level lifetime, charge state distribution, and average charge of iron plasmas.     

Effects of thermal aging on Fe ion-irradiated Fe–0.6%Cu alloy investigated by positron annihilation

Thermal aging effects on surface of 2.5 Me V Fe ion-irradiated Fe–0.6%Cu alloy were investigated using positron annihilation techniques. The samples were irradiated at 573 K to a dose of 0.1 dpa. Their thermal aging was performed at 573 K for 5, 50 and 100 h. From the results of Doppler broadening measurement, an obvious trough could be seen in near-surface region from the S parameters and inflection point form at S–W curves. This indicates changes in the annihilation mechanism of positrons in surface region after thermal aging. Coincident Doppler broadening indicates that the density of Cu precipitates in the thermal aged samples decreased, due to recovery of the vacancies.     

Influence of angle’s ranges for recording an X-ray fluorescence hologram on reconstructed atomic images

1 Introduction cal beam paths can be reversed and positions of the source and detector can be exchanged. This method Solid structures are usually determined by means can be regarded as the inverse of XFH and is calledof X-ray diffraction. However, the diffraction tech- “inside detector holography”. The former method isnique suf…