Novel α + β Zr Alloys with Enhanced Strength

Low-alloyed zirconium alloys are widely used in nuclear applications due to their low neutron absorption cross-section. These alloys, however, suffer from limited strength. Well-established guidelines for the development of Ti alloys were applied to design new two-phase ternary Zr alloys with improved mechanical properties. Zr-4Sn-4Nb and Zr-8Sn-4Nb alloys have been manufactured by vacuum arc melting, thermo-mechanically processed by annealing, forging, and aging to various microstructural conditions and thoroughly characterized. Detailed Scanning electron microscopy (SEM) analysis showed that the microstructural response of the alloys is rather similar to alpha + beta Ti alloys. Duplex microstructure containing primary alpha phase particles surrounded by lamellar alpha + beta microstructure can be achieved by thermal processing. Mechanical properties strongly depend on the previous treatment. Ultimate tensile strength exceeding 700 MPa was achieved exceeding the strength of commercial Zr alloys for nuclear applications by more than 50%. Such an improvement in strength more than compensates for the increased neutron absorption cross-section. This study aims to exploit the potential of alpha + beta Zr alloys for nuclear applications.

Boron-Based Neutron Scintillator Screens for Neutron Imaging

In digital neutron imaging, the neutron scintillator screen is a limiting factor of spatial resolution and neutron capture efficiency and must be improved to enhance the capabilities of digital neutron imaging systems. Commonly used neutron scintillators are based on 6LiF and gadolinium oxysulfide neutron converters. This work explores boron-based neutron scintillators because 10B has a neutron absorption cross-section four times greater than 6Li, less energetic daughter products than Gd and 6Li, and lower γ-ray sensitivity than Gd. These factors all suggest that, although borated neutron scintillators may not produce as much light as 6Li-based screens, they may offer improved neutron statistics and spatial resolution. This work conducts a parametric study to determine the effects of various boron neutron converters, scintillator and converter particle sizes, converter-to-scintillator mix ratio, substrate materials, and sensor construction on image quality. The best performing boron-based scintillator screens demonstrated an improvement in neutron detection efficiency when compared with a common 6LiF/ZnS scintillator, with a 125% increase in thermal neutron detection efficiency and 67% increase in epithermal neutron detection efficiency. The spatial resolution of high-resolution borated scintillators was measured, and the neutron tomography of a test object was successfully performed using some of the boron-based screens that exhibited the highest spatial resolution. For some applications, boron-based scintillators can be utilized to increase the performance of a digital neutron imaging system by reducing acquisition times and improving neutron statistics.

Safety features of fast reactor with heavy atomic weight weakly neutron absorbing reflector

The purpose of the present study is the justification of the possibility of improving fast reactor safety by surrounding reactor cores with reflectors made of material with special neutron physics properties. Such properties of 208Pb lead isotope as heavy atomic weight, small neutron absorption cross section, and high inelastic scattering threshold result in certain peculiarities in neutron kinetics of the fast reactor equipped with 208Pb reflector, which can significantly enhance reactor safety. The reflector will also make possible generation of additional delayed neutrons characterized by the “dead” time. This will improve the resistibility of the fission chain reaction to stepwise reactivity excursions and exclude prompt supercriticality. Let us note that generation of additional delayed neutrons can be shaped by reactor designers. The relevance of the study amounts to the fact that generation of additional delayed neutrons in the reflector will make it possible mitigating the consequences of a reactivity accident even if the introduced reactivity exceeds the effective fraction of delayed neutrons. At the same time, the role of the fraction of delayed neutrons as the maximum permissible reactivity for reactor safety is depreciated. Scientific originality of the study pertains to the fact that the problem of yield of additional neutrons with properties close to normal delayed neutrons, has not been posed before. The authors suggest a new method for enhancing safety of fast reactors by increasing the fraction of delayed neutrons due to the time delay of prompt neutrons during their transfer in the reflector. In order to benefit from the expected advantages, the following combination is acceptable: lead enriched by 208Pb is used as a neutron reflector while natural lead or other material (sodium, etc.) is used as a coolant in the reactor core.

Actinide and Lanthanide Adsorption onto Hierarchically Porous Carbons Beads: A High Surface Affinity for Pu

Structured carbon adsorbents were prepared by carbonizing macroporous polyacrylonitrile beads whose pores were lined with a mesoporous phenolic resin. After activation, the beads were tested for minor actinide (Np and Am), major actinide (Pu and U) and lanthanide (Gd) adsorption in varying acidic media. The activation of the carbon with ammonium persulfate increased the surface adsorption of the actinides, while decreasing lanthanide adsorption. These beads had a pH region where Pu could be selectively extracted. Pu is one of the longest lived, abundant and most radiotoxic components of spent nuclear fuel and thus, there is an urgent need to increase its security of storage. As carbon has a low neutron absorption cross-section, these beads present an affordable, efficient and safe means for Pu separation from nuclear waste.

Influence of Xe-135 Dynamic Behavior on Core Operation Safety for a Molten Salt Reactor

The isotope Xe-135 has a large thermal neutron absorption cross section and is considered to be the most important fission product. A very small amount of such neutron poison may significantly affect the reactor reactivity since they will absorb the neutrons from chain reaction, therefore monitoring their atomic density variation during reactor operation is extremely important. In a molten reactor system, Xe-135 is entrained in the liquid fuel and continuously circulates through the core where the neutron irradiation functions and the external core where only nuclei decay occurs, at the same time, an off-gas removal system operates for online removing Xe-135 through helium bubbling. These unique features of MSR complicate the Xe-135 dynamic behaviors, and the calculation method applied in the solid fuel reactor system is not suitable. From this point, we firstly analytically deduce the nuclei evolution law of Xe-135 in the flowing salt with an off-gas removal system functioning. A study of Xe-135 dynamic behavior with the core power change, shutdown, helium bubbling failure and startup then is conducted, and several valuable conclusions are obtained for MSR design.

Solving Bateman Equation for Xenon Transient Analysis Using Numerical Methods

After a nuclear reactor is shutdown, xenon-135, an isotope with a very high thermal neutron absorption cross-section, will build up and reduce the reactivity considerably for a while. This is known as poisoning. However, the concentration xenon-135 would gradually decrease through decaying or absorbing neutron, making it necessary to suppress the reactivity in order to prevent the reactor to go critical or supercritical. Thus, it is important to predict the relationship between xenon poisoning and time after the reactor is shutdown to ensure the safety of the reactor. This paper reports the research on the prediction of xenon poisoning in a hypothetical nuclear reactor after it is shut down. In order to make the prediction, the Bateman equations of xenon (Xe) and iodine (I), which is of the form of an Ordinary Differential Equation (ODE) system, need to be solved. Two different methods, the fourth-order Runge-Kutta method and the matrix exponential method, were applied to solve the ODE system with MATLAB codes. The accuracies and computational efficiencies of the two method is also studied and compared.

Gadolinium tracers for enhancement of Sigma-log contrast measurements

Borehole neutron measurements are routinely used for in situ rock assessment in hydrocarbon reservoirs. We have used gadolinium oxide nanoparticles for enhancing the sensitivity of macroscopic thermal neutron absorption cross-section (Sigma) measurements of rocks. The gadolinium-based doping agent is used due to its exceptionally high neutron absorption cross section, low cost, and availability; it is also shown to preserve or enhance the differentiation between pore fluids. Injected from a pilot well, the doping agent could thus substantially improve the precision of Sigma-derived saturation measurements. Computational modeling verifies that modest gadolinium concentrations in the rock’s pore volume give rise to significant enhancement of reported Sigma: A concentration of 1750 weight ppm is shown to offer superior contrast enhancement across a wide range of rock solid and fluid compositions. Preliminary experimental work confirms the reported effects obtained with computational modeling of gadolinium doping in simulated porous rocks.