Low Neutron Absorption Cross Section Research Articles

Microstructure and Properties of Pressureless-Sintered Zirconium Carbide Ceramics with MoSi2 Addition.

Zirconium carbide (ZrC) ceramics have a high melting point, low neutron absorption cross section, and excellent resistance to the impact of fission products and are considered to be one of the best candidate materials for fourth-generation nuclear energy systems. ZrC ceramics with a high relative density of 99.1% were successfully prepared via pressureless sintering using a small amount of MoSi2 as an additive. The influence of the MoSi2 content on the densification behavior, microstructure, mechanical properties, and thermal properties of ZrC ceramics was systematically investigated. The results show that the densification of ZrC was significantly enhanced by the introduction of MoSi2 due to the formation of a liquid phase during sintering. In addition, the ZrC grains were refined due to the pinning effect of the generated silicon carbide. The flexural strength and Vickers hardness of ZrC ceramics with 2.5 vol% MoSi2 sintered at 1850 °C were 408 ± 12 MPa and 17.1 GPa, respectively, which were approximately 30% and 10% higher compared to the samples without the addition of MoSi2. The improved mechanical properties were mainly attributed to the high relative density (99.1%) and refined microstructure.

Hot deformation behaviour of Swaged Zircaloy-4 at varying strain rate and its microstructural evolution

Zirconium alloy demonstrates a favourable combination of strength, ductility, low neutron absorption cross-section, as well as excellent corrosion and oxidation resistance. In the context of its application as a reactor component, the critical hot deformation behaviour of isothermally swaged Zr-4 alloy has been extensively studied. Different combinations of elevated temperatures (873 K to 1023 K) and strain rates (0.01 s−1 to 10/s) have shown enhancement in flow stress. Analysis of stress flow, microstructural changes, and activation energy behaviour was performed using a state-of-the-art thermomechanical simulator with 50% reduction in sample thickness. The modified Arrhenius equation was employed to correlate theoretical and experimental flow stress behaviour. EBSD was utilized to understand texture evolution in hot deformed samples, revealing a transformation from pure basal to basal plus pyramidal poles concerning the compression direction. Texture evolution played a significant role in microstructural development, further characterized through TEM, which showed influential characteristics like subgrain boundaries, dislocations, and precipitate behaviour. Dynamic recrystallization mechanisms in swaged Zr-4 hot deformed samples were analysed using bright and dark field TEM modes, which were subsequently correlated with other microstructural characteristics such as grain morphology and structural defects.

Tribological and corrosion behavior of hydrided zirconium alloy

Zirconium alloys (Zr-alloys) are used for the core components of nuclear reactors due to their exceptional properties, which make them suitable for the given applications such as low neutron absorption cross-section, good corrosion resistance, good wear resistance, good mechanical stability at high temperatures etc. The coolant water that flows inside the pressure tube contains LiOH to control the pH value of the coolant water, which can cause corrosion. The flow of coolant water produces low amplitude vibration, which can cause fretting wear of the components. Further, the coolant water at high temperature causes the hydride formation in the material, degrading the properties of zirconium alloys. The safe working of nuclear reactors is the prime objective, and the corrosion and wear behavior of the core components must be explored to prevent possible failure. Fretting wear or tribo-corrosion experiments were conducted on Zr‐2.5Nb alloy against SS-410 for varied fretting duration to understand the tribological response. Electrochemical corrosion experiments were performed under the aqueous solution of LiOH (10.4 pH) using Electrochemical Impedance Spectroscopy (EIS) and potentiodynamic polarization technique. The comparison was made between the tribological and corrosion response of unhydrided‐Zr‐2.5Nb alloy (Zr‐2.5Nb) and hydrided- Zr‐2.5Nb alloy (Zr‐2.5Nb‐H). The hydride formation causes the reduction in corrosion resistance of Zr‐2.5Nb alloy. The oxide layer is formed during the wear and corrosion process which protects the material from further wear and corrosion respectively.

Numerical analysis of multi-field coupling behaviors of delayed hydride cracking for thin plate-type Zr-2.5Nb specimens

Zirconium alloys are extensively utilized as structural materials in nuclear applications due to their exceptional mechanical strength, corrosion resistance, and low thermal neutron absorption cross-section. The stress concentration provoked by the defects in zirconium alloy components will initiate delayed hydride cracking (DHC), significantly jeopardizing nuclear safety. In the present investigation, an enhanced cohesive model that incorporates the effects of stress state has been proposed to realize the three-dimensional simulation of multi-field coupled DHC behaviors in thin plate-type zirconium alloy specimens. The simulation results indicate that: (1) the consistence of the predicted DHC velocities with the corresponding experimental data validates the newly developed cohesive model; (2) the strip characteristics presented in the cracked surface are captured, demonstrating that the varying fracture patterns across the thickness are well simulated; (3) the stress-state induced hydrostatic-stress gradients and solid-soluted hydrogen concentration gradients make against the formation of a hydride strip to induce hydride embrittlement near the plate surfaces, resulting in a slower cracking of the surface layers than that of the internal region; for the thinner specimens the asynchronous cracking phenomena through the thickness will greatly affect the multi-field coupling behaviors around the mid-plane; the non-cracked surface layers will apply the dragging effects to the new cracking of the internal region, leading to longer waiting time and thicker hydrided zone before fracture.

Microstructural and mechanical property evolution of a nuclear zirconium-4 alloy fabricated via laser powder bed fusion and annealing heat treatment

ABSTRACT Zirconium (Zr) alloys are widely used in nuclear energy because of their excellent mechanical properties and low thermal neutron absorption cross-section. This work investigated the printability, microstructure, and mechanical properties of Zr-4 alloy additively manufactured by laser powder bed fusion (LPBF) for the first time. The effect of annealing temperature on the microstructural and the mechanical property evolution of the printed Zr-4 alloy was studied. The results exhibited that the Zr-4 alloy with a high relative density of 99.77% was obtained using optimised printing parameters. With an increase in the annealing temperature, the formed α phase of the Zr-4 alloy changed from an acicular shape to a coarse-twisted shape, and finally to an equiaxed shape. Such microstructure change endowed the alloy with a high compressive strength of 2130 MPa and compressive strain of 36%. When the annealing temperature exceeded 700°C, Zr x (Fe2Cr) compounds were precipitated, strengthening the alloy by pinning effect. These findings provide valuable guidance for the manufacture of geometrically complex Zr alloy parts for nuclear power applications.

VNbCrMo refractory high-entropy alloy for nuclear applications

Refractory high-entropy alloys (RHEAs) with high melting points and low neutron absorption cross-section are sought for generation-IV fission and fusion reactors. A high throughput computational screening tool, Alloy Search and Predict (ASAP), was used to identify promising RHEA candidates from over 1 million four-element equimolar combinations. The selected VNbCrMo RHEA was further studied by CALPHAD to predict phase formation, which was compared to an experimentally produced ingot aged at 1200 °C. The VNbCrMo RHEA was found to constitute a majority bcc phase, with a 6% area fraction of C15-Laves formed at interdendritic regions, in contrast to the predictions of single-phase. The prediction of the yield strength by a model based upon edge dislocation mechanisms indicated 2.1 GPa at room temperature and 850 MPa at 1000 °C for the equimolar single bcc phase. The hardness of the alloy with C15-Laves was 748 HV (yield strength ∼2.4 GPa). Finally, the macroscopic neutron absorption cross-section was modelled for a wide range of energies. Displacements per atom per year and activation calculations, up to 1000 years after 2 years of continuous operation, in typical fusion and fission reactor scenarios were also performed using the inventory code FISPACT-II. This work gives new insight into the phase stability and performance of the VNbCrMo RHEA, which is compared with a similar design concept alloy, to assess the potential of novel RHEAs for use in advanced nuclear applications.

Grain Size Control by Matrix Phase Additives in Nuclear Thermal Propulsion Cermets

Refined grain sizes in the metallic phase of a cermet nuclear fuel have been reported to provide improved performance and survivability of the fuel in nuclear thermal propulsion (NTP) applications. In this paper, we study the use of ZrC nanoparticles, which have a low neutron absorption cross-section, as Zener pinning sites for the tungsten matrix in a surrogate fuel - yttria stabilized zirconia (YSZ) – cermet. While the ZrC additions were found to successfully pin grain boundaries, subsequent high temperature hydrogen exposure at 2500 K for 1 hour resulted in significantly more catastrophic cracking in comparison to control specimens with no ZrC additions. This increase in cracking is explained by two gas formation mechanisms - hydrogen recombination at defect sites and water vapor formation from the reduction of YSZ.

Preparation, microstructure and ion-irradiation damage behavior of Al4SiC4-added SiC ceramics

Single-phase Al4SiC4 powder with a low neutron absorption cross section was synthesized and mixed with SiC powder to fabricate highly densified SiC ceramics by hot pressing. The densification of SiC ceramics was greatly improved by the decomposition of Al4SiC4 and the formation of aluminosilicate liquid phase during the sintering process. The resulting SiC ceramics were composed of fine equiaxed grains with an average grain size of 2.0 μm and exhibited excellent mechanical properties in terms of a high flexure strength of 593 ± 55 MPa and a fracture toughness of 6.9 ± 0.2 MPa m1/2. Furthermore, the ion-irradiation damage in SiC ceramics was investigated by irradiating with 1.2 MeV Si5+ ions at 650 °C using a fluence of 1.1 × 1016 ions/cm2, which corresponds to 6.3 displacements per atom (dpa). The evolution of the microstructure was investigated by X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy. The breaking of Si–C bonds and the segregation of C elements on the irradiated surface was revealed by XPS, whereas the formation of Si–Si and C–C homonuclear bonds within the Si–C network of SiC grains was detected by Raman spectroscopy.

Joining of Zirconium Alloys to Nickel Bearing Alloys for In-Core Components

Abstract Zirconium alloys are well-positioned as the material of choice for nuclear in-core structures since they have a low neutron absorption cross section and possess good material properties under elevated neutron flux and high-temperature operating conditions. The Canadian supercritical water-cooled reactor (SCWR) concept pressure tube includes a transition from zirconium-tin alloy (Excel) to a yet to be selected stainless steel to allow it to be seal welded to the inlet plenum. Thus, a method to join zirconium and its alloys to ferrous and nickel-bearing alloys is desired. To this end, a series of experiments were carried out to demonstrate the feasibility of this method. A joint approaching the strength of the parent material was achieved using both rotary friction welding and co-extrusion methods, although the joints exhibited lower ductility than that of the parent material. Microstructural examination of the joint revealed a ∼200 nm to ∼1 μm intermetallic layer at the transition zone, which was the location of failure in the tensile specimen.

Evaluation of thermophysical properties of UO2-10 Vol% Mo nuclear fuel pellets

Uranium dioxide (UO 2 )−molybdenum (Mo) composites have been proposed as accident tolerant fuel candidates due to their high thermal conductivity , high melting point , high boiling point, and ability to have low thermal neutron absorption cross-section. It is therefore essential to evaluate the method of fabrication and thermophysical properties of this system. In this work, UO 2 -Mo samples with a Mo fraction of 10 Vol% and different densities were fabricated using spark plasma sintering (SPS). This method allows sintering at relatively low temperatures and within a short time. Experimental analysis was performed on UO 2 -Mo specimens to characterize thermal diffusivity and thermal conductivity as a function of temperature up to 900 °C. Thermophysical properties were analyzed for various densities, and grain sizes as a function of temperature. The grain size and microstructural differences related to porosity were also analyzed using the scanning electron microscopic (SEM) and electron backscatter diffraction (EBSD).

Irradiation Hardening and Microstructure Characterization of Zr -1% Nb During Low Dose Neutron Irradiation

Due to low neutron absorption cross section, high mechanical strength, high thermal conductivity and good corrosion resistance in water and steam, Zirconium alloys are widely used as fuel cladding material in nuclear reactors. During life-time of a reactor the microstructure of this alloy is affected due to, among other factors, radiation damage and hydrogen damage. In this work mechanical properties changes on neutron irradiated Zr-1wt.% Nb at low temperatures (< 100 °C) and low dose (3.5 ´ 1023 n m-2 (E > 1 MeV)) were correlated with hydrides and crystal defects evolution during irradiation. To achieve this propose, tensile tests of: 1) Non-hydrided and non-irradiated material, 2) Hydrided and non-irradiated material and 3) Hydrided and irradiated material were performed at 25 ºC and 300 ºC. Different phases, hydrides and second phase precipitates were characterized by transmission electron microscopy (TEM) techniques. For the hydrided and irradiated material, the ductility decreased sharply with respect to the hydrided and non-irradiated material, among other factors, due to the change in the microstructure produced mainly by neutron irradiation. Even if the presence of the hydride ζ (zeta) was observed, both in the irradiated and non-irradiated material, tensile tests showed that ζ-hydrides did not affect ductility, since hydrided samples are more ductile than non-hydrided samples.

Forced Convection Heat Transfer of Zirconia-water Nanofluid in Vertical Triangular Sub-channel

Recently, the researchers in nuclear technology are thinking about the use of nanofluid as an alternative cooler, besides using the base-fluids or water and gas. The idea of using nanofluid is already being considered to be applied in nuclear power plants in the future. This nanofluid is expected to be applied in cooling systems, both primary cooling systems and emergency core cooling systems. Therefore, this study was conducted to determine the characteristics of the thermo-physical properties of ZrO2-water nanofluid in the forced convection flow. Recent studies showed that ZrO2-water nanofluid has a good prospect to be used in the nuclear reactor technology due to its low neutron absorption cross section. Although several papers have reported the physical properties of ZrO2-water nanofluid, practically there is no correlation equation for predicting forced convection heat transfer in a vertical triangular sub-channel in ZrO2-water nanofluid. This experiment used ZrO2–water nanofluid at a constant nanofluid concentration of 0.10 wt.%. The results showed that the average increase in ZrO2-water nanofluid heat transfer coefficient was greater than 20% compared to the base-fluids or water heat transfer coefficient as coolant. It is expected that this nanofluid will support the current trend especially in cooling system applications.

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.

Synthesis and Characterization of a Carborane-containing Precursor for B<sub>4</sub>C Ceramics

B₄C ceramic is one of widely used ceramics with properties of low density, high hardness, high thermal-stability and high neutron absorption cross-section. However, the high melting point and high hardness of B₄C make it difficult to be prepared as materials with complex shape. Precursor derived method is an effective approach to prepare special-shaped ceramics. In the early research of B₄C ceramic precursors, highly toxic and unstable decaborane was generally used as the boron source material, and precursors were also unstable and difficult to synthesize. In this work, a new B₄C precursor is designed and synthesized, and m-carborane, which has stable chemical properties and high thermal stability, is used as the boron source. Firstly, m-carborane is converted into lithium salt after removing active hydrogen by n-butyllithium. Then, catalyzed by Cu(I), the lithium salt reacts with m-chloroiodobenzene so that monomer, bis(3-chlorophenyl)-m-carborane, is prepared. The monomer is polymerized through Ullmann Coupling reaction at the present of Ni(0) catalyst and Zinc powder, and precursor poly[bis(3-chlorophenyl)-m-carborane] is finally synthesized. The precursor is stable at room temperature in air and soluble in Tetrahydrofuran, with Mw of 1013g/mol, and ceramic yield of 62.20% at 900°C in N₂. The precursor pyrolytic product is composed of B₄C and graphite after heat treated at 1600°C in inert atmosphere.

Finite Element Analysis and Experimental Evaluation of Residual Stress of Zr-4 alloys Processed through Swaging

Zirconium alloy has been extensively used as a cladding material in nuclear power reactors due to its low neutron absorption cross section, excellent mechanical properties, and corrosion resistance. The influence of the swaging parameter, feed rate (0.7, 1.25, 2 m/min) on residual stress induced in Zr-4 alloy is investigated in the present work. A three-dimensional finite element model was implemented in the Deform 3D software to simulate the rotary swaging (RS) process over a circular rod of Zr-4 alloy. The simulation results based on the 3D framework provide a detailed insight of residual stress, true stress versus true strain and force applied over the rod during the multiple pass swaging process; the results are compared with experimental results. The experimental hole drilling method is used to determine the residual stresses on swaged zirconium alloy at different feed rates (0.7, 1.25, and 2 m/min). A similar trend of residual stress between experimental and numerical results from the surface to the center on the swaged rod samples is observed. The same magnitude of residual stress at the surface of the swaged Zr-4 rod is also observed. It is found to be compressive at the surface and tensile in the center of the samples, as observed in the present work.

Neutronic Analysis of Lead208-Bismuth Eutectic-Cooled Modified CANDLE Reactor with Core Geometry Variations

Lead208-Bismuth Eutectic (LBE)- cooled reactor is one of the Generation IV reactors. The fourth-generation has several objectives: economic competitiveness, inherent safety, minimize radioactive waste, and non-proliferation. Modified CANDLE burnup scheme is one solution to fulfil these objectives. The scheme does not need fuel enrichment and reprocessing Pu from LWR. Lead208-Bi eutectic was employed as a coolant of the reactor due to its low neutron absorption cross-section. However, However, it is necessary for surveying the neutronic core design. In this study, we compared the core neutronic among the variations of core geometry with equal volume of Pb208-Bi eutectic-cooled Modified CANDLE reactor. The calculation is performed by PIJ and CITATION of SRACs module. The Pancake cylindrical (larger diameter) has the highest K-eff and the lowest-peaking factor among of balance and tall cylindrical.

Corrosion analysis of post-heat treatment and post-weld SS316 with electrokinetic reactivation and cyclic polarization method

Austenitic stainless steel 316 has very high mechanical properties and corrosion resistance. This type of steel is widely used both in the nuclear and non-nuclear industries. In the nuclear industry, SS316 is used as a cladding material for uranium fuel due to its good corrosion and mechanical properties, and also low neutron absorption cross-section. In the Center for Nuclear Fuel Technology (PTBBN BATAN), it is used as material for the container of nuclear waste that is to be stored on Temporary Storage Installation of Spent Fuel (KHIPSB3 BATAN). SS316 is used as material for can to contain high-activity solid waste from the testing activity in Radiometallurgy Installation (IRM BATAN). The lid of the container is sealed with the GTAW welding process in order to tightly contain the solid waste. The main problem with the heat treatment and welding process of austenitic stainless steel is the occurrence of sensitization in a temperature range of 500-800°C. Therefore fully electrochemical analysis of SS316 stainless steel in various mediums has been conducted. SS 316 specimen was heat-treated to simulate the heat generated by the welding process. Sensitization analysis was conducted with qualitative and quantitative methods by EIS and EPR, and pitting corrosion resistance was by cyclic polarization method. The solution used for EIS and cyclic polarization test was 0 − 3.5% concentration of NaCl, and for EPR test was a mixture of H2SO4 and KSCN. Material characterization before and after corrosion testing was microstructure examination. The result of the corrosion test showed that heat treatment on the temperature range of 500-800°C caused sensitization. The corrosion test curve result showed that a heat treatment temperature of 650°C for 1 hour had the highest activation current. The lowest Rp value for SS316 specimen post heat treatment in 675°C was 69.410 ohm. The welded SS316 specimen had a higher corrosion current than that of the unwelded specimen. The microstructure of the welded specimen showed that there was intergranular corrosion particularly in the HAZ region. The effect of NaCl concentration on the cyclic polarization test showed that the higher the NaCl concentration, the more easily the pitting corrosion. The indication of pitting corrosion occurrence was evaluated by considering the Epit and Erp values. The lower Epit value meant that pitting corrosion was more easily to occur.

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.

Structural characterization of Zircaloy-4 subjected to helium ions irradiation of variable fluence

Zircaloy-4 is widely used as structural material for the fuel cladding of nuclear reactors due to its low neutron absorption cross-section, excellent corrosion resistance and good mechanical stability under neutron irradiation. In nuclear reactors, helium ions (He+) are produced during (n, α) reactions that can affect the structural performance of the Zircaloy-4. Therefore, it is important to investigate the structural properties of Zircaloy-4 after irradiation with helium ions. In this work, structural properties of He+irradiated Zircaloy-4 are investigated at different ion fluences. The samples of Zircaloy-4 were irradiated with 300 keV He+ at different fluences (1 × 1013, 1 × 1015 and 1 × 1016 ions/cm2) by using Pelletron Accelerator at room temperature. The ion range inside the samples was 0.84 µm as estimated through the SRIM software. Structural properties of the pristine and He+irradiated samples were studied through x-ray diffraction that revealed polycrystalline Zr peaks whose intensity and width were changed after irradiation. The diffraction peaks of Zr corresponding to (002) plane were found to be higher in intensity as compared to its other diffraction peaks. The changes in peaks positions at lower and higher diffraction angles were also noticed that was attributed to the generation of stresses inside the irradiated samples. The Williamson-Hall (W-H) analysis of the samples depicted a decrease in material's crystallinity after exposure to He+ up to 1 × 1015 ions/cm2 whereas the crystallinity was started to improve with increase of the ion fluence to 1 × 1016 ions/cm2. These results were explained on the basis of collisional cascades process and defects annihilation under the thermal spike model.

Studies on texture and formability of Zircaloy-4 produced by pilgering route

Zircaloy-4 is a widely used material in nuclear industries due to its excellent corrosion resistance at elevated temperatures, low neutron absorption cross section and balanced strength and ductility. In the present study, the formability of Zircaloy-4 sheets produced through pilgering route has been investigated. Nakajima dome tests have been conducted on sheet samples taken along rolling, transverse and diagonal directions with respect to the rolling direction of the sheet to construct the forming limit diagrams (FLDs). The strain signature of deformed samples has also been studied to determine the peak limiting strains. Further, limiting dome height values are calculated and all the results are validated with FE-Simulations. Finally, XRD studies have been performed to understand the relationship between the texture and formability. The FLD and texture studies reveal that the Zircaloy-4 possesses higher drawability in comparison to that of the stretchability. The orientation of the specimen has negligible influence on FLD in tension–tension region and marginal influence in tension–compression region.