•Thermodynamic equilibrium state calculations have been conducted to assess the oxidation and corrosion reactions of various neutron absorber materials in various environments such as oxygen, steam, ...air, air–steam mixtures.•It is explored the potential of oxide-based neutron absorber compounds formed by combining neutron absorber oxide materials with highly oxidation-resistant materials.•The oxide-based neutron absorber compounds are expected to extend the lifetime of the LWR control rod as well as improve stability and resistance to oxidation and corrosion.
The challenges associated with conventional boron carbide (B4C) control rod materials, including helium gas accumulation and susceptibility to oxidation and corrosion in various environments, have been thoroughly explored. To address these issues, a comprehensive investigation into the potential of oxide-based neutron absorber compounds for control rods has been undertaken. Thermodynamic equilibrium state calculations have been conducted to assess the oxidation and corrosion reactions of various neutron absorber materials (Gd, Hf, Sm, Er, Eu, Dy-based oxide), including B4C, in various environments such as oxygen, steam, air, and air–steam mixtures. The results have unveiled vulnerabilities of B4C in these environments, notably the generation of gases such as H2, CO, and boron compounds. Furthermore, neutron absorber oxide materials have exhibited potential susceptibility to oxidation and corrosion in steam environments. Consequently, the potential of oxide-based neutron absorber compounds, formed by combining neutron absorber oxide materials with highly oxidation-resistant substances (ZrO2, TiO2), has been explored. Thermodynamic equilibrium state calculations indicate that these compounds maintain robust resistance to oxidation and corrosion across various environments. This paper demonstrates the superiority of oxide-based neutron absorber compounds as alternatives to existing boron carbide neutron absorber materials. Additionally, the oxide-based neutron absorber compounds are expected to extend the lifetime of the LWR control rod as well as improve stability and resistance to oxidation and corrosion.
Accelerated corrosion tests of Al-B4C neutron absorber, equivalent to 37 months in a spent nuclear fuel pool, were conducted on three different period-installed surveillance coupons (33, 52, and 92 ...months) to further investigate the underlying mechanisms of premature surface corrosion and 10B depletion which we had recently reported in another study. Microstructure characterization, electrochemical analysis, and neutron attenuation tests were conducted after the corrosion tests, and two types of galvanic corrosion, Al matrix/stainless steel and Al matrix/B4C particles, were discovered. The duplex oxide layer comprised of amorphous oxide and Al(OH)3 films and pit corrosion were also observed on the surface with reduced densities (< 1%). The neutron attenuation tests on the absorbers and ICP-MS measurement of the coolant showed no considerable change in 10B areal density.
A novel (10 wt%B4Cp+3.6 wt%Gd)/Al6061 neutron absorber material having great potential commercial applications was designed by calculating equivalent B content (BEq) and its neutron absorber ability ...was evaluated based on an equivalent B areal density (EBAD) calculation as well as a Monte Carlo simulation. The designed material was successfully fabricated by ultrasound assisted casting method. The added B4C particles were distributed uniformly in the matrix and the newly formed large-sized Al3Gd particles existed along the grain boundaries (GBs) in the as-cast composite. It was found that a small amount of Si was solubilized in Al3Gd lattice and the solution behavior of Si was revealed using first-principles calculation. After hot extrusion (HE) and heat treatment (HT), the large-sized Al3Gd particles were broken into small ones and nano-sized βʺ phase was precipitated in the matrix. The mechanical properties of the modified composite were enhanced remarkably and the reason of which was mainly attributed to the following two aspects. On the one hand, HE induced the grain refinement and the fragmentation of large-sized Al3Gd particles as well as their more homogeneous distribution within grains from GBs were beneficial for the improvements in both strength and ductility of composite. On the other hand, HT induced the precipitation of βʺ phase could work as strengthening phase in the modified composite. The size and distribution of Al3Gd particles played an important role in improving the mechanical properties since cracking easily occurred on the large-sized Al3Gd particles which existed along GBs, leading to the severe degradation of mechanical properties of the as-cast composite. Furthermore, the related mechanism of cracking behavior of large-sized Al3Gd particles was discussed. This research provides a low-cost method to prepare easy-deform Al based neutron absorber material with desirable mechanical properties.
Boron carbide is a strategic material, finding applications in nuclear industry, armour for personnel and vehicle safety, rocket propellant, etc. Its high hardness makes it suitable for grinding and ...cutting tools, ceramic bearing, wire drawing dies, etc. Boron carbide is commercially produced either by carbothermic reduction of boric acid in electric furnaces or by magnesiothermy in presence of carbon. Since many specialty applications of boron carbide require dense bodies, its densification is of great importance. Hot pressing and hot isostatic pressing are the main processes employed for densification. In the recent past, various researchers have made attempts to improve the existing methods and also invent new processes for synthesis and consolidation of boron carbide. All the techniques on synthesis and consolidation of boron carbide are discussed in detail and critically reviewed.
Rare earth compounds (RE2O3), typically Eu2O3 and Gd2O3, are considered as absorbing materials due to their long chain neutron absorbing ability and high efficiency of neutron absorption. In this ...paper, a specifically designed xEu2O3-(1-x)Gd2Zr2O7 (x = 0.6, 0.7, 0.8, 0.85) composite ceramics with excellent thermal properties were synthesized through direct thermal decomposition reactions followed a high-temperature sintering method under vacuum conditions. X-ray diffraction (XRD) and Raman spectroscopy analysis indicated that the obtained ceramics consisted of monoclinic Eu2O3 and fluorite Gd2Zr2O7 phases. The hardness of xEu2O3-(1-x)Gd2Zr2O7 ceramics (x = 0.8, 0.85) have no significantly changes in the test temperature range (25–350 °C). The thermal expansion coefficients (TECs) (7.3–10 ×10-6 K-1) are lower than Dy2TiO5 (9.3–10.4 ×10-6 K-1), and the thermal expansion rate changes linearly with the increase of temperature, which indicates that the xEu2O3-(1-x)Gd2Zr2O7 ceramics exhibit excellent high temperature phase stability. Furthermore, the range of thermal conductivities is from 0.96 W·m-1·K-1 to 2.02 W·m-1·K-1 (25–800 °C), which is higher than Dy2TiO5 (0.27–0.55 W/m·K, 220–650 °C). Compared with the state-of-art Dy2TiO5 ceramics, xEu2O3-(1-x)Gd2Zr2O7 have superior thermal properties and are therefore expected to be the next generational of neutron absorbent materials used in ash rod.
Al-B4C neutron absorbers in spent fuel pools have been assumed to have negligible radiation damage and helium generation from 10B(n, α)7Li reactions. However, surveillance coupons have shown highly ...radiation-damaged microstructure. We conducted 200 keV He+ ion irradiation on Al 6061 at three different doses (0.1, 1, and 10 dpa) and compared the microstructures to the surveillance coupon. The neutron absorber's bubble size (25.9 ± 7.4 nm) was most similar to the 10 dpa irradiated specimen (28.4 ± 13.8 nm). Commercial neutron absorber (BORAL and MAXUS®) were also irradiated up to 10 dpa and showed preferential cavity formation along grain boundaries.
Higher enrichment of nuclear fuel along the manufacturing limit of boron content in steel and aluminum alloys represents a significant challenge in designing spent fuel transport and storage ...facilities. One possible solution for spent fuel pools and casks is the burnup credit method that allows for decreasing very high safety margins associated with fresh fuel assumption in spent fuel facilities. An alternative solution based on placing neutron absorber material directly into the fuel assembly is proposed here. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the assemblies. The efficiency of the newly proposed concept is demonstrated on the criticality safety analysis of the GBC-32 spent fuel cask. Absorber rods from 8 different elements are placed within Westinghouse OFA 17x17 guide tubes. Currently used boron is a good option because of high absorption cross section, low atomic mass and chemical compatibility with various alloys. Alternative options (e.g., Sm, Eu, Gd, Dy, Hf, Re, Ir) are based on very good absorbers that do not require alloy compatibility since the absorbers can be placed inside zirconium or steel cladding. Because of high efficiency of the newly proposed absorber concept, boron content in BORAL sheets can be decreased to more competitive economics. Moreover, fuel assembly pitch is investigated in order to change cask wall inner diameter that will result in lower material consumption for the cask wall with the same shielding thickness.
•Analyzed of neutron characteristics and material damage on neutron absorber by the irradiation test using HANARO research reactor.•Clarified the model specification of the HANARO reactor core and ...irradiation system.•Described the characteristics of boron carbide neutron absorber and its nuclear characteristics.•Explained the correlation between the nuclear absorption characteristics and material damage of neutron absorption materials.
This study aims to verify the nuclear characteristics and material integrity of neutron absorbers through neutron irradiation using the HANARO research reactor. The nuclear reactions and material damage of boron carbide in the neutron irradiation environment of the external region of the reactor core were evaluated based on the Monte Carlo method. The irradiation conditions of the reactor core were assumed based on equilibrium core operation with a reactor power of 30 MW. In order to evaluate material damage over a one-year period, it was assumed that the concentrated boron carbide undergoes a 2 % boron depletion every 100 days. Reactivity test was assessed to verify the appropriateness of the assumptions established for the irradiation test. As a result, the reactivity level showed approximately 1.06 mk, which sufficiently proves its lower level compared to the criteria of 12.5 mk. In order to analyze the physical characteristics of neutron absorbers influencing material damage, the group fluxes and energy-dependent neutron spectra of the target material were evaluated. The neutron flux in Group I appears low, while the epithermal group flux is observed relatively high. The specimen material closer to the core generally showed higher flux, indicating a maximum difference of about 3 × 1013 #/cm2·sec compared to the minimum one. The 3rd specimen showed the highest in flux, while 7th specimen is the lowest. The shape of all the dpa values showed similar trends to the pattern of neutron absorption reaction rates. The neutron absorption was significant up to 100 days, and after this point, they showed gradual reductions in the damage increase rate. The total dpa values linearly increased along with the irradiation time, and showed about 12–13 dpa over one year of EFPD. The results of this study can be utilized as fundamental data for confirming changes in material damage due to high neutron irradiation on neutron absorbers in connection to nuclear characteristics.
Recent experimental reports on the premature corrosion of an Al-B4C neutron absorber in a spent nuclear fuel pool, which was possibly assisted by 10B(n, α)7Li reaction-induced porous microstructures, ...illustrated the need for quantitative evaluation of the neutron-induced energetic particle emission reactions inside neutron absorbers. For this purpose, we developed a Safety Analysis code for NeuTron Absorbers in spent nuclear fuel pools (SANTA) by integrating several existing codes (TRITON, ORIGEN, CSAS6, and modified SDTrimSP) with newly developed modules to provide the essential parameters for the experimental emulation of irradiation-assisted corrosion of the absorbers. The most important outputs of this code are radiation damage in displacement per atom (dpa) units and He concentration in atomic ppm. These outputs are required for the design of heavy-ion irradiation experiments on the absorbers to emulate the radiation damage induced by energetic 7Li ions and α-particles. The SANTA code is also user-friendly: its whole sequence can be executed using simplified text-based input without user intervention. This code may improve the understanding of the gas bubble formation and irradiation-assisted corrosion mechanisms of the neutron absorbers, and it could provide a foundation for the future development of more robust safety codes for spent nuclear fuel pools.
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