Effect of Gd on neutron absorption properties and electrochemical corrosion behavior of Zr-Gd alloy in boiling concentrated HNO₃

Effect of Gd on neutron absorption properties and electrochemical corrosion behavior of Zr-Gd alloy in boiling concentrated HNO 3 1 2 first-author 2 corresp xiaoganghu@yeah.net xiaoganghu@yeah.net xiaoganghu@yeah.net https://orcid.org/0000-0002-7459-6105 1 corresp pyguo@just.edu.cn pyguo@just.edu.cn pyguo@just.edu.cn 2 2 School of Materials Science and Engineering, Jiangsu University of Science and Technology, Zhenjiang 212100, China School of Materials Science and Engineering, Jiangsu University of Science and Technology, Zhenjiang 212100, China Xi'an Rare Metal Materials Institute Co., Ltd, Xian 710016, China Xi'an Rare Metal Materials Institute Co., Ltd, Xian 710016, China A Zr-Gd alloy with neutron poisoning properties and resistance to boiling concentrated HNO3 corrosion was developed based on a corrosion-resistant Zr-702 alloy to meet the demand for neutron shielding in the closed-loop treatment of spent fuel and the nuclear chemical industry. In this study, 1 wt. %, 3 wt. %, 5 wt. %, 7 wt. %, and 9 wt. % Zr-Gd alloys were designed and fabricated with Zr-702 as the control element. The electrochemical behavior of the Zr-Gd alloys in boiling concentrated HNO3 was investigated, and the neutron shielding effect on plate thickness and Gd content was simulated. The experimental results demonstrate that the corrosion resistance of the alloy decreased slightly before ~7 wt. %–9 wt. % with increasing Gd content; this is the inflection point of its corrosion resistance. The alloy uniformly dissolved the Gd content that could not be dissolved in the Zr lattice, resulting in numerous micropores on the passivation coating, which deteriorated and accelerated the corrosion rate. The MCNP simulation demonstrated that when the Gd content was increased to 5 wt. %, a 2-mm-thick plate can shield 99.9 % neutrons; an alloy with a Gd content ≥7 wt. % required only a 1-mm-thick plate, thereby showing that the addition of Gd provides an excellent neutron poisoning effect. Thus, the corrosion resistance and neutron shielding performance of the Zr-Gd alloy can meet the harsh service requirements of the nuclear industry. Zr-Gd alloy Boiling concentrated HNO3 Electrochemistry Neutron shielding MCNP Introduction 1 As the number of reactors being built and in operation increases, the generation and accumulation of spent fuel will result in potential radiation risks. There are two main treatment protocols for the spent fuel. The closed-loop treatment of offloaded spent fuel is more effective than traditional deep-burial treatment. The realization of a nuclear fuel closed-loop cycle reduces the pollution of nuclear waste and improves the utilization rate of Actinide element materials. The main post-treatment technology is to use boiling concentrated HNO3 to dissolve the spent fuel, and then the Purex process is used to extract uranium and plutonium (for recycling) from the spent fuel by an organic extractant [ 1 1 , 2 2 ] . Therefore, the container materials used in spent-fuel reprocessing need to be robust under the harsh environment of high temperature, strong acid, and intense irradiation. The consequences of corrosion penetration are wide ranging, therefore alloys (with neutron absorption properties) that can be used at high temperatures and display a resistance to corrosion by concentrated HNO3 have become necessary materials for key equipment used during and after spent fuel treatment. Adding neutron poisonous elements to some corrosion-resistant alloy systems is expected to result in good neutron absorption properties, excellent mechanical properties, and corrosion resistance. Currently, nuclear-class stainless steel 316 L, Ti-Ta alloy Ti35, and Zr alloys have been extensively adopted for use in the nuclear industry [ 3 3 - 11 11 3-11 ] . Boron is frequently added to alloys or other shielding materials as the preferred neutron poison element owing to its low cost and good neutron absorption capacity. Boron stainless steel (BSS) has excellent mechanical properties [ 12 12 - 14 14 12-14 ] ; however, it exhibits severe intracrystalline corrosion in concentrated HNO3 solutions containing oxidation ions [ 15 15 , 16 16 ] . Even more severe intercrystalline cracking can occur in a high-radiation environment [ 17 17 - 19 19 17-19 ] . Boron-titanium alloys exhibit excellent wear resistance and strength. However, the plasticity displays a marked reduction under irradiation [ 20 20 ] , and the alloy displays even the loss of malleability in severe cases [ 21 21 ] . The alloy also swells owing to the transmutation of B, similar to BSS [ 22 22 ] . It appears that the addition of non-metallic boron to the alloy is self-limiting. As for zirconium and titanium alloys, in contrast to zirconium alloys, titanium alloys appear to exhibit an evident condensate corrosion phenomenon [ 23 23 ] . At high temperatures, the corrosion resistance of zirconium alloys is preferable to that of titanium alloys. Mattern et al. [ 24 24 , 25 25 ] reported the relationship between the solubility of Gd in Ti and Ti in Gd, and the temperature based on Vegard's law. The solubility of Gd in body centered cubic BCC-Ti at 1800 K was ~0.6–0.3 at. %, and the maximum solubility of Ti in BCC-Gd is ~0.8–0.3 at. %. Combined with the phase diagram of Gd-Ti and the above results, it can be seen that the solubility of Gd and Ti in the corresponding solid alloys is minimal. According to the Zr-Gd phase diagram [ 26 26 ] , the solubility of Zr in the HCP-Gd base phase is 3.8 at.% (6.38 wt. %) at 1160 ℃, and the solubility of Zr in the HCP-Gd base phase decreases with a decrease in temperature. The maximum solubility of Gd in a BCC-Zr base phase appears to be approximately 1500 ℃, which is approximately 6.5 at. %, and then decreases with increasing or decreasing temperature. The maximum solubility of Gd in the HCP Zr phase was 2.8 at. % (4.73 wt. %) at 880 ℃, and the solubility decreases with a decrease in temperature. In addition, the solid alloy system is in a mixed crystal state of Gd and Zr. Therefore, zirconium alloys, which exhibit excellent corrosion resistance, good mechanical properties [ 27 27 ] , and superior thermal conductivity [ 28 28 ] , have become the preferred materials for nuclear chemical equipment. Zirconium alloys do not possess neutron absorption capacity; therefore, it is necessary to add neutron toxic elements to improve the neutron shielding performance. However, the corrosion resistance of boron–zirconium alloys becomes markedly worse, and alloy swelling occurs during irradiation [ 29 29 ] . Gadolinium is another element, which may be considered. Gadolinium has the highest neutron shielding capability performance and can be added to zirconium alloys as a neutron poison. There are a few reasons for selecting gadolinium. Firstly, natural gadolinium has two isotopes, 155 Gd-60600b and 157 Gd-139000b, with extremely high neutron absorption cross-sections, which are considerably higher than that of 10 B-3837b. Secondly, under long-term corrosion conditions, the degradation rate of gadolinium in water is lower than that of the borides [ 30 30 , 31 31 ] . However, studies on the addition of the neutron toxicant gadolinium to zirconium alloys are still lacking. This study aims to explore the influence of adding gadolinium to zirconium alloys on their corrosion resistance and neutron properties. In this study, Zr-Gd alloys with different Gd contents were prepared by non-consumable arc melting. To simulate the actual service conditions, 6 mol/L concentrated HNO3 at 95 ℃ was used for the test. The relationship between the Gd content in the Zr-Gd alloy and the corresponding corrosion resistance properties was studied. Neutronic calculations of the shielding of the Zr-Gd alloy were performed using the MCNPX-2.7.0 Monte Carlo method. Experiments 1 Preparation of Zr-Gd alloy 2 A vacuum non-consumable arc melting furnace was employed to smelt 150 g of ingot castings. Varying amounts (1 wt. %, 3 wt. %, 5 wt. %, 7 wt. %, and 9 wt. %) of gadolinium were added to zirconium sponge; zirconium sponge (Zr-702) was used as a control. The zirconium sponge was mixed with the gadolinium chips and placed in a crucible. To obtain uniform Zr-Gd alloy ingots, the melting current was set at 300–500 A; each ingot was flipped and melted six times. The Zr-Gd alloy ingot underwent solid solution (1050 ℃, 1 h) treatment, quenching, hot rolling (800 ℃ insulation for 30 min, 20–50% of channel pressure), annealing (500 ℃; 45 min), and scale removal. Finally, the plate was machined to the required size. Electrochemical corrosion test 2 The Zr-Gd alloy plates were cut into 12.5 mm×9.5 mm×1.5 mm-size samples. The surfaces and cross-sections of the samples were polished with water abrasive papers of 60 #, 240 #, 600 #, 600 #, 1200 #, and 2000 #, successively, to avoid gap corrosion during electrochemical corrosion. Zirconium wire was used as a conductive wire to avoid concentrated HNO3 infiltration into the plates. Electrochemical performance tests were performed using a Koster CS350M Electrochemical Workstation (Wuhan, China). The Zr-Gd alloy was used as the working electrode, the reference electrode was a saturated calomel electrode (Ag/AgCl saturated KCl solution), and the counter electrode was a platinum electrode placed in the three-electrode system electrolyte. A schematic diagram of the device is shown in Fig. 1 Fig. 1 . The solution temperature was maintained at 95±1 ℃. The corrosion medium was a 6 mol/L concentrated HNO3 solution. Zr-Gd alloys with different Gd concentrations were placed in this system for dynamic potential scanning measurements. Finally, the polarization curve and alternating current impedance spectrum were obtained. The electrochemical parameters were set as follows: scan from the initial potential of -1 V to the termination potential of 3 V, at a scanning rate of 1 mV/s. The open-circuit potential measurement occurs before the electrochemical impedance spectroscopy (EIS) measurement. The sample was immersed in the electrolyte for approximately 3,600 s until its open-circuit potential was stable. The scanning frequency ranged from 100 kHz to 0.1 Hz, with a sinusoidal perturbation potential amplitude of 10 mV. The surface morphology was observed using a desktop SEM (COXEM EM30AX+, 12 kV, Korea) after corrosion. The laser intensity data were reconstructed using a 3D laser micro-imaging system (VK-150K, Kez, Japan), and the surface 3D images and height distributions of the samples were obtained. To detect the chemical compositions on the Zr-Gd alloy surface, XPS (Thermo Fischer, ESCALAB 250Xi) analysis was performed with a monochromatic Al Kα ( hv = 1486.6 eV) source under a pressure of 4×l0 -9 mbar, with a working voltage of 14.6 kV, a filament current of 13.5 mA, and signal accumulation for 20 cycles. The test pass-energy was 20 eV, the step size was 0.1 eV, and the charge correction was performed with a C 1s = 284.8 eV combined energy standard. Results and discussion 1 Microstructure analyses 2 Fig. 2 Fig. 2 shows the X-ray diffraction patterns of the Zr-Gd alloys with different Gd concentrations. The acronym hcp indicates a close-packed hexagonal crystal structure. The numbers in brackets indicate the crystal indexes. Zr-702 mainly exists in the α-Zr phase with an HCP crystal structure; the (002) and (101) lattice planes grow in competition. The overall left-shift of the crystal plane is due to the addition of the solid solution of Gd atoms in the zirconium lattice. With the addition of gadolinium, the gadolinium signal gradually increases but is not obvious. The Gd (102) plane is evident at 42.6°, wherein the existence of the hcp phase is completely demonstrated. However, the content is low because the main phase Zr signal is so strong that it overwhelms the Gd peak. Up to 7 wt.% or higher, (100), (002), and (101) can be found at 28.8°, 31.137°, and 32.411°, respectively. Gadolinium is insoluble in the zirconium HCP lattice at low temperatures; it can only melt into Zr at approximately 3.8 at.% (6.38 wt.%) at 1160 ℃. Although quenching retains the solution state at high temperatures, it also precipitates gadolinium atoms. Therefore, gadolinium phase precipitation can be clearly observed when the content is above 7 wt. %. Metallographic corrosion photos and EDS surface scanning were prepared to determine the distribution of gadolinium within the HCP structure in the Zr-Gd alloy, as shown in Fig. 3 Fig. 3 . Metallography ( Fig. 3a-1-d-1 Fig. 3a-1-d-1 ) shows that the grain size increases significantly in the Zr-Gd alloy with an increase in the content of gadolinium. As gadolinium is added, it is uniformly dissolved in the zirconium lattice, thus inhibiting the nucleation rate of the Zr metal. Gadolinium increases the viscosity of the alloy liquid, reduces the condenser depression of the zirconium alloy, and provides more crystallization time for the grain growth of the zirconium alloy. EDS mapping analysis ( Fig. 3a-3-d-3 Fig. 3a-3-d-3 ), indicates that HCP Gd is evenly distributed in the grains of the zirconium alloy. When the concentration is less than 7 wt. %, the gadolinium is uniformly distributed in the grains and no visible precipitation occurs. However, when the content is ≥7 wt. %, uniform precipitation occurs while the gadolinium is uniformly distributed in the grains, but not concentrated at the grain boundaries. Open-circuit potential 2 The open-circuit potential is the electrode potential at zero current density, and the positive and negative values of the electrode potential determine the tendency of the alloy corrosion. The electrode potential of the alloy in the electrolyte gradually stabilized. The gadolinium content was inversely proportional to the open-circuit potential, and the addition of gadolinium decreased the open-circuit potential. As shown in Fig. 4 Fig. 4 , when the electrode potential is stable, the electrode potentials of Zr, 1 wt. %, and 3 wt. % are 0.98 V, 0.96 V, and 0.93 V, respectively. The electrode potential gradually decreases, indicating that the corrosion tendency gradually increases. However, the decrease in the electrode potential is very low, and its corrosion tendency is the same. When the gadolinium content in the alloy reaches 5 wt. % and 7 wt. %, the electrode potential is maintained at the 0.85 V horizontal line, proving that the corrosion tendency under this composition is constant. The electrode potential of 9 wt. % is 0.75 V, and is more pronounced, so the component Zr-Gd alloy is the most prone to corrosion. It can be assumed that the 7 wt. % alloy composition belongs to the inflection point of the alloy. This may be because under the high-temperature solution condition of 1050 ℃, the HCP-Gd atom can maintain approximately 6 wt. % content in α-Zr (HCP) and fully enter the zirconium lattice. The gadolinium atom is fixed in the α-Zr phase by cold quenching, forming a composite crystal structure of the α-Zr (Gd) and β-Zr phases. Excess Gd atoms beyond the solution range precipitate and become a preferential point for corrosion. Polarization curve and corrosion morphology 2 Fig. 5 Fig. 5 shows the corrosion polarization curves of Zr-Gd alloys with different Gd concentrations at 95 ℃ (6 mol/L HNO3). The polarization curve is divided into four potential zones: the active dissolution zone, transition passivation zone, stable passivation zone, and over-passivation zone, where the point corresponding to the peak is the critical passivation point. When the curve exceeds the critical passivation region, the current density increases with the potential. This shows that the corrosion and regeneration rate of the passivation film reaches dynamic equilibrium at this time. The passivation zones corresponding to Zr (1 wt. %, 3 wt. % and, 5 wt. %) are stable at the corrosion current density of 1×10 -5 A/cm 2 . The corrosion current density of 7 wt. % is slightly higher than those of the other concentrations, indicating that the corrosion rate increases slightly but is generally consistent, while the corrosion current density of 9 wt.% rises to 1×10 -4 A/cm 2 . The passivation films of Zr (1 wt. %, 3 wt. %, 5 wt. % and 7 wt. %) are more stable and have better protective performance than the passivation films of the 9 wt. % alloy when the Zr-Gd alloy is formed. This is related to the maximum solubility of gadolinium in zirconium; gadolinium destroys the corrosion resistance of zirconium and gadolinium alloys when its content exceeds a reasonable level (about 6.38 wt. %). The critical potential of the destruction of the passivation film of the Zr-Gd alloy is approximately 1.75 V. When the potential exceeds 1.75 V at high temperature, the corrosion current density of the Zr-Gd alloy rises sharply into the over-passivation zone. The passivation film is destroyed and ruptured because the passivation film generated by the zirconium alloy in concentrated HNO 3 or high-temperature HNO 3 is porous and unstable [ 32 32 , 33 33 ] . The corrosion potential and current density can be calculated in the transition zone using the Tafel method [ 34 34 ] , as shown in Table 1 Table 1 . Fig. 6 Fig. 6 shows the variation diagram of the corrosion potential and current density of Zr-Gd alloys with different gadolinium concentrations in a 6 mol/L HNO3 (95 ℃) solution. With an increase in the gadolinium concentration, the corrosion current density of the Zr-Gd alloy gradually increases, and positively correlates with the gadolinium concentration. The corrosion current density of the 9 wt. % alloy rises sharply compared with that of the other concentrations, once again, confirming that the 7 wt. % alloy composition is the corrosion inflection point. However, the corrosion potential is inversely correlated with the gadolinium concentration. This tendency for corrosion is consistent with the results of the opening circuit potential. The stability of the Zr-Gd alloy oxidation film worsens and it is prone to damage and spot corrosion [ 35 35 ] . Spot corrosion undergoes self-catalysis and accelerates the reaction, and the dissolution rate in the hole is expected to be significant, resulting in accidents [ 36 36 , 37 37 ] . 1 1 Alloy 1 1 E corr (mV) 1 1 I corr (A/cm²) 1 1 Zr 1 1 981.6 1 1 1.04×10 –7 1 1 1 wt. % 1 1 976.5 1 1 5.23×10 –7 1 1 3 wt. % 1 1 945.8 1 1 4.01×10 –6 1 1 5 wt. % 1 1 903.3 1 1 8.82×10 –6 1 1 7 wt. % 1 1 854.7 1 1 1.18×10 –5 1 1 9 wt. % 1 1 740.0 1 1 1.75×10 –4 This is shown in Fig. 7 Fig. 7 and Fig. 8 Fig. 8 . The residual passivation film on the alloy surface has many small holes and severe peeling marks after the polarization treatment. It is indicated that the passivation film generated on Zr-Gd alloy surface is an unstable porous passivation film. When the voltage exceeds the critical voltage, the passivation film is damaged and peels off because the corrosion rate is greater than the regeneration rate. The alloy corrosion rate accelerates and the corrosion current density increases sharply. As shown in Fig. 8 Fig. 8 , the surface of the 5 wt. % Zr-Gd alloy produces evenly distributed micro porous spot corrosion. This is caused by the corrosion and shedding of Gd atoms or precipitated atomic groups. The evenly distributed shedding has a destructive effect on the oxide film. The corrosion liquid can enter and pass through the oxide film to continue the ion exchange, producing a continuous corrosion effect. The sample with a concentration of 9 wt. % (5.43 at. %) performs poorly in terms of its mechanical properties and total immersion corrosion; the micropores become more extensive, and signs of deterioration occur in overlapping sheets. Although not detached, it can be assumed that an extension has occurred internally. Corrosion is accelerated by continuously destroying the alloy surface via perforation corrosion. To determine the specific composition of the residual oxide film generated on the surface after electrochemical corrosion, the surface composition was tested using XPS and the curve was calibrated. Gd was detected using 1 wt. %–9 wt. % through full spectrum analysis. To fit and divide the peaks of the O element, three peaks must correspond, as shown in Fig. 9 Fig. 9 . The Zr-Gd alloy mainly contains Zr and Hf from Zr-702 and also contains 1–9 wt. % Gd. According to a comparison with the NIST XPS database, the binding energies of ZrO 2 , Gd 2 O 3 , and HfO 2 are ~529.90–531.30 [ 38 38 , 39 39 ] , ~529.5–531.4, [ 40 40 ] and ~530.0–530.40 [ 41 41 ] , respectively. The residual oxide film was determined to be a mixture of ZrO 2 , Gd 2 O 3 , and HfO 2 . With an increase in the Gd content, the proportion of the integral area of ZrO 2 to the that of the oxide film decreases, which proves that the dense oxide film is gradually corroded and penetrated, and reduces the corrosion resistance of the alloy. The three-dimensional morphology characterization of the 3 wt.%, 5 wt.%, and 7 wt.% alloys were tested by laser microscopy, as shown in Fig. 10 Fig. 10 . Sdr is the interface expansion area ratio; the surface roughness increases as Sdr increases. Fig. 10 Fig. 10 shows the three-dimensional topography of the specified area with a magnification of 200×. where Sdr is the surface roughness of the shot area. With an increase in the gadolinium content, the Sdr value of the 3 wt.%, 5 wt.%, and 7 wt.% alloys are 0.6477, 1.194, and 1.748, respectively; the Sdr values of the Zr-Gd alloys increase gradually after polarization. Fig. 10 Fig. 10 shows that the surface of the Zr-Gd alloy exhibits uniform corrosion and low roughness. With the increase in the gadolinium content, the stability of the passivation film produced by the Zr-Gd alloy also gradually decreases. Combined with the surface morphology, a local range of cracking and accelerated corrosion is observed. Alternating current impedance spectrum of the Zr-Gd alloy 2 Fig. 11 Fig. 11 shows the Nyquist and locally amplified plots of the Zr-Gd alloy with different gadolinium concentrations in a 6 mol/L HNO3 solution at 95 ℃. The radius of the capacitive arc represents the corrosion resistance of an alloy in a corrosive medium. The larger the radius, the better is the corrosion resistance of the alloy. Zirconium has the maximum arc-resistance radius, indicating the best corrosion resistance. The capacitive arc radius gradually decreases with an increasing gadolinium content, indicating that the corrosion resistance gradually decreases. This is consistent with the experimental results of the polarization curve. Fig. 12 Fig. 12 shows the Bode plots of the Zr-Gd alloy in high-temperature-concentrated HNO3. Fig. 12(a) Fig. 12(a) shows the change curve of the impedance amplitude and disturbance frequency. The larger the mold value, the better is the corrosion resistance of the alloy. With an increase in the gadolinium content, the mode value of the Zr (1 wt. %, 3 wt. %, 5 wt. %, and 7 wt. %) alloys gradually decreases to a minimum, and its corrosion resistance does not mean much difference. The 9 wt. % component alloy has a mode value that is significantly lower than those of the alloys comprising the other gadolinium concentrations, which agrees with the experimental results of the polarization curve. Fig. 12(b) Fig. 12(b) shows the change curve of the relationship between the phase angle and the disturbance frequency. The larger the phase angle and the wider the curve, the better the corrosion resistance of the alloy. The phase angle and frequency relationship curve are divided into two arcs, and the second arc of the 1 wt. %, 3 wt. %, 5 wt. %, and 7 wt.% alloys is roughly the same size. The alloy arc of the 9 wt. % Zr alloy is significantly smaller than those of the other alloys containing differing concentrations of gadolinium. Fig. 13 Fig. 13 shows the equivalent circuit-diagram model used for the EIS analysis. The equivalent circuit can be described as a resistor and constant phase angle element. The equivalent impedance spectral circuit of the Zr-Gd alloy in 6 mol/L HNO3 (95 ℃) solution fitted the corrosion potential by using ZSimpWin software tool. The impedance spectra provided the best equivalent fitting to the circuit R (C(R(CR))). Here, R s is the solution resistance, representing the electrode impedance between the reference electrode and working electrode, R por is the oxide film resistance generated by the alloy, R ct is the charge transfer resistance of the electrode during the charge transfer process, and CPE is a constant phase angle element (understood as a capacitor) formed on the electric bilayer capacitor surface between the working electrode and corrosion medium. The continuous phase-angle element CPE expression is as follows: [ 42 42 ] <math display="block" id="M1"><mrow><msub><mi>Z</mi><mrow><mtext>CPE</mtext></mrow></msub><mrow><mo>(</mo><mi>w</mi><mo>)</mo></mrow><mo>=</mo><msup><mrow><mrow><mo>[</mo><mrow><mi>C</mi><msup><mrow><mrow><mo>(</mo><mrow><mi>j</mi><mi>w</mi></mrow><mo>)</mo></mrow></mrow><mi>n</mi></msup></mrow><mo>]</mo></mrow></mrow><mrow><mo>−</mo><mn>1</mn></mrow></msup><mo>,</mo></mrow></math> where n is the dispersion index (-1≤ n ≤1), when n =1, the constant phase angle element indicates pure capacitance; when n = 0, the continuous phase angle element represents pure resistance, ω represents the frequency, and c represents the capacitance. Fig. 13 Fig. 13 shows that the polarization resistance is R p = R por + R ct . The electrochemical parameters of Zr-Gd alloys in 6 mol/L HNO3 (95 ℃) solution were obtained using an equivalent circuit, as shown in Table 2 Table 2 . Because the corrosion resistance of the Zr alloy is realized by the generated passivation film ZrO 2 , the corrosion resistance of the alloy can be determined by the polarization resistance, R p . With an increase in the Gd content, the polarization resistance and R ct of the alloy gradually decrease. It has been proven that with the addition of excess gadolinium, the Zr-Gd alloy passivation film will be porous, unstable, can burst, and even fall off. The difficulty in ion exchange between the passivation films is notably reduced. The charge transfer resistance R ct decreases, resulting in the transport of ions, and the charge of the corrosion resistance is reduced, accelerating the passivation film failure. As the corrosion rate of the alloy increases, the corrosion resistance of the alloy decreases. The corrosion resistance of Zr-702 is the same as that of 1 wt.%, 3 wt.%, 5 wt.%, and 7 wt.%, and a sudden drop occurs at 9 wt.%, which is also consistent with the results expressed by the polarization curve. 1 1 Alloying constituent 1 1 R s (Ω cm -2 ) 1 1 CPE por (μFcm -2 ) 1 1 n 1 1 1 R por (Ω cm -2 ) 1 1 CPE dl (μFcm -2 ) 1 1 n 2 1 1 R ct (Ω cm -2 ) 1 1 Zr 1 1 1.176 1 1 5.573 1 1 0.98 1 1 1134 1 1 37.07 1 1 0.82 1 1 10,974 1 1 1 wt. % 1 1 1.029 1 1 6.779 1 1 0.96 1 1 1588 1 1 41.17 1 1 0.82 1 1 8059 1 1 3 wt. % 1 1 1.022 1 1 4.157 1 1 0.99 1 1 2783 1 1 64.66 1 1 0.80 1 1 4695 1 1 5 wt. % 1 1 0.992 1 1 5.481 1 1 0.97 1 1 2590 1 1 74.84 1 1 0.73 1 1 4049 1 1 7 wt. % 1 1 1.244 1 1 7.854 1 1 0.96 1 1 1883 1 1 88.92 1 1 0.75 1 1 2758 1 1 9 wt. % 1 1 0.896 1 1 16.05 1 1 0.92 1 1 1546 1 1 66.33 1 1 0.93 1 1 2288 Thermal neutron absorption properties of Zr-Gd alloy 2 The composition and thickness of the shielding material affected the shielding performance. The Monte Carlo simulation method is currently a widely used critical and shielding calculation method. The MCNP is a universal geometric, time-correlated, combined neutron-photon-electron transport Monte Carlo particle transport program. It can create continuous/multi-group energy cross-sections and solve the problems of neutrons, photons, electronic transportation, and joint transportation. The Monte Carlo model requires the density of the Zr-Gd alloy. The theoretical density of metallic zirconium is 6.49 g/cm³, and that of metallic gadolinium is 7.901 g/cm³. The theoretical density of the Zr-Gd alloy for each component is listed in Table 3 Table 3 . 1 1   1 1 at% 1 1 wt.% 1 1 ρ (g/cm 3 ) 1 1 1 wt. % 1 1 Gd 0.58 Zr 99.42 1 1 Gd 1 Zr 99 1 1 6.501611 1 1 3 wt. % 1 1 Gd 1.76 Zr 98.24 1 1 Gd 3 Zr 97 1 1 6.524958 1 1 5 wt. % 1 1 Gd 2.96 Zr 97.04 1 1 Gd 5 Zr 95 1 1 6.548473 1 1 7 wt. % 1 1 Gd 4.18 Zr 95.82 1 1 Gd 7 Zr 93 1 1 6.572158 1 1 9 wt. % 1 1 Gd 5.43 Zr 94.57 1 1 Gd 9 Zr 91 1 1 6.596015 1 1 11 wt. % 1 1 Gd 6.69 Zr 93.31 1 1 Gd 11 Zr 89 1 1 6.620047 Fig. 14 Fig. 14 simulates the thermal neutron transmittance, x , of Zr-Gd alloys with different Gd concentrations, where the shielding rate (absorption rate) is 1- x . The Gd concentration calculated by neutron shielding simulation was extended to a broader range, that is, 1 wt. % to 19 wt. %. According to the local magnification diagram, when the thickness of the Zr-Gd plate is 1 mm, the 1 wt.% Zr-Gd alloy will pass through 31.5 % thermal neutrons, to shield 99.9 % thermal neutrons; the thickness of the plate must be 7 mm, more and thicker Zr-Gd alloy material are needed to achieve the neutron absorption effect of 99.9%; therefore, the shielding effect is not good. When the Gd content is increased to 3 wt. %, the alloy can reduce the neutron transmittance to 3 % by a 1 mm plate, and the decreasing effect is significant; a 3 mm plate can absorb 99.9 % thermal neutrons. A 2 mm plate can completely shield neutrons when the concentration is 5 wt. %, and a 1 mm plate thickness can completely shield neutrons when the concentration is 7 wt. % or more. Conclusion 1 In this study, Zr with excellent corrosion resistance and Gd with good neutron absorption performance were selected and developed to obtain Zr-Gd alloys for monitoring the corrosion resistance and neutron shielding function. The following conclusions are drawn. (1) The corrosion resistance of the Zr-Gd alloy gradually decreases with increasing Gd content. (2) When the Gd content in the alloy was ≤7 wt.%, the corrosion resistance of the alloy remained almost constant when the Gd content in the alloy was ≤7 wt.%, although there was a negligible decreaseThe corrosion resistance rapidly decreases between 7 wt.% and 9 wt.%, proving that ~7–8 wt.% is the inflection point of the alloy corrosion resistance. (3) A porous and unstable passivation film was formed on the Zr-Gd alloy by applying an overpass voltage in boiling concentrated HNO 3 . (4) When the Gd content was 3 wt.%, 99.9% of the hot neutrons can be absorbed without the need for a plate of 2 mm thickness. With an increase in the Gd content, the absorption effect of the hot neutrons becomes more apparent. (5) The thermal neutrons are absorbed after elastic collision with Gd atoms, and produce secondary gamma-ray hazards [ 43 43 , 44 44 ] . In future studies, the neutron absorption and secondary gamma-ray absorption properties of Zr-Gd alloys must be considered for optimization by adding heavy elements, such as W, Ta, and Bi. References J. E. Birkett , M. J. Carrott , O.D. Fox et al. , Recent developments in the Purex process for nuclear fuel reprocessing: Complexant based stripping for uranium/plutonium separation . CHIMIA. Int. J. Chem 59 , 898 - 904 ( 2005 ). doi: 10.2533/000942905777675327 http://doi.org/10.2533/000942905777675327 . Recent developments in the Purex process for nuclear fuel reprocessing: Complexant based stripping for uranium/plutonium separation R.S. Herbst , P. Baron , M. Nilsson . Standard and advanced separation: PUREX processes for nuclear fuel reprocessing . Adv. Sep. Tech. Nucl Fuel. Reprocess , 141 - 75 ( 2011 ). doi: 10.1533/9780857092274.2.141 http://doi.org/10.1533/9780857092274.2.141 . Standard and advanced separation: PUREX processes for nuclear fuel reprocessing P. Deng , Q. Peng , E.-H. Han , et al. , Effect of irradiation on corrosion of 304 nuclear grade stainless steel in simulated PWR primary water . Corros. Sci 127 , 91 - 100 ( 2017 ). doi: 10.1016/j.corsci.2017.08.010 http://doi.org/10.1016/j.corsci.2017.08.010 . Effect of irradiation on corrosion of 304 nuclear grade stainless steel in simulated PWR primary water F. Hua , K. Mon , P. Pasupathi et al. , A review of corrosion of titanium grade 7 and other titanium alloys in nuclear waste repository environments . Corros 61 , 987 - 1003 ( 2005 ). doi: 10.5006/1.3280899 http://doi.org/10.5006/1.3280899 . A review of corrosion of titanium grade 7 and other titanium alloys in nuclear waste repository environments U.K. Mudali , R. Dayal , J. Gnanamoorthy , Corrosion studies on materials of construction for spent nuclear fuel reprocessing plant equipment . J. Nucl. Mater. 203 , 73 - 82 ( 1993 ). doi: 10.1016/0022-3115(93)90432-X http://doi.org/10.1016/0022-3115(93)90432-X . Corrosion studies on materials of construction for spent nuclear fuel reprocessing plant equipment U. K. Mudali , A.R. Shankar , R. Natarajan et al. , Application of zirconium alloys for reprocessing plant components . Nucl. Tech 182 , 349 - 57 ( 2013 ). doi: 10.13182/NT12-73 http://doi.org/10.13182/NT12-73 . Application of zirconium alloys for reprocessing plant components D.O. Northwood . The development and applications of zirconium alloys . Mater. Des 6 , 58 - 70 ( 1985 ). doi: 10.1016/0261-3069(85)90165-7 http://doi.org/10.1016/0261-3069(85)90165-7 . The development and applications of zirconium alloys R. Rajasekaran , A.K. Lakshminarayanan , M. Vasudevan et al. , Role of welding processes on microstructure and mechanical properties of nuclear grade stainless steel joints . Proc. Inst. Mech. Eng, Part L 233 , 2335 - 51 ( 2019 ). doi: 10.1177/1464420719849448 http://doi.org/10.1177/1464420719849448 . Role of welding processes on microstructure and mechanical properties of nuclear grade stainless steel joints D.W. Shoesmith , J. Noel , D. Hardie et al. , Hydrogen absorption and the lifetime performance of titanium nuclear waste containers . Corros. Rev 18 , 331 - 60 ( 2000 ). doi: 10.1515/CORRREV.2000.18.4-5.331 http://doi.org/10.1515/CORRREV.2000.18.4-5.331 . Hydrogen absorption and the lifetime performance of titanium nuclear waste containers S. Wang , M. Zhang , H. Wu et al. , Study on the dynamic recrystallization model and mechanism of nuclear grade 316LN austenitic stainless steel . Mater. Charact 118 , 92 - 101 ( 2016 ). doi: 10.1016/j.matchar.2016.05.015 http://doi.org/10.1016/j.matchar.2016.05.015 . Study on the dynamic recrystallization model and mechanism of nuclear grade 316LN austenitic stainless steel Y. Han , J. Mei , Q. Peng et al. , Effect of electropolishing on corrosion of nuclear grade 316L stainless steel in deaerated high temperature water. Corros . Sci: J. Environ. Degrad. Mater. Control 112 , ( 2016 ). doi: 10.1016/j.corsci.2016.09.002 http://doi.org/10.1016/j.corsci.2016.09.002 . Effect of electropolishing on corrosion of nuclear grade 316L stainless steel in deaerated high temperature water. Corros J. Jiménez , P. Adeva , M.C. Cristina et al. , Characterization of rapidly solidified ultrahigh boron steels . Mater. Sci. Eng. A 159 , 103 - 109 ( 1992 ). doi: 10.1016/0921-5093(92)90403-N http://doi.org/10.1016/0921-5093(92)90403-N . Characterization of rapidly solidified ultrahigh boron steels J. Martin , Effects of processing and microstructure on the mechanical properties of boron-containing austenitic stainless steels . Waste. Proc ( 1989 ). doi: 10.1016/0921-5093(95)10001-6 http://doi.org/10.1016/0921-5093(95)10001-6 . Effects of processing and microstructure on the mechanical properties of boron-containing austenitic stainless steels P. Acosta , J.A. Jimenez , G. Frommeyer et al. , Microstructural characterization of an ultrahigh carbon and boron tool steel processed by different routes . Mater. Sci. Eng. A 206 , 194 - 200 ( 1996 ). doi: 10.1016/0921-5093(95)10001-6 http://doi.org/10.1016/0921-5093(95)10001-6 . Microstructural characterization of an ultrahigh carbon and boron tool steel processed by different routes S. Ningshen , U.K. Mudali , S. Ramya et al. , Corrosion behaviour of AISI type 304L stainless steel in nitric acid media containing oxidizing species . Corros. Sci 53 , 64 - 70 ( 2011 ). doi: 10.1016/j.corsci.2010.09.023 http://doi.org/10.1016/j.corsci.2010.09.023 Corrosion behaviour of AISI type 304L stainless steel in nitric acid media containing oxidizing species R. Priya , S. Ningshen , U.K. Mudali , Influence of oxidizing ion concentration on the corrosion resistance of type 304L stainless steel in nitric acid medium . Corros 69 , 335 - 44 ( 2013 ). doi: 10.5006/703 http://doi.org/10.5006/703 . Influence of oxidizing ion concentration on the corrosion resistance of type 304L stainless steel in nitric acid medium L.P. Andresen , Irradiation Assisted Stress Corrosion Cracking . Compr. Nucl. Mater , 177 - 205 ( 2012 ). doi: 10.1016/B978-0-08-056033-5.00084-7 http://doi.org/10.1016/B978-0-08-056033-5.00084-7 . Irradiation Assisted Stress Corrosion Cracking K. Fukuya , M. Nakano , K. Fujii et al. , IASCC susceptibility and slow tensile properties of highly-irradiated 316 stainless steels . J. Nucl. Sci. Technol 41 , 673 - 81 ( 2004 ). doi: 10.1080/18811248.2004.9715532 http://doi.org/10.1080/18811248.2004.9715532 . IASCC susceptibility and slow tensile properties of highly-irradiated 316 stainless steels H. Nishioka , K. Fukuya , K. Fujii et al. , IASCC Initiation in Highly Irradiated Stainless Steels under Uniaxial Constant Load Conditions . J. Nucl. Sci. Technol 45 , 1072 - 7 ( 2008 ). doi: 10.1080/18811248.2008.9711894 http://doi.org/10.1080/18811248.2008.9711894 . IASCC Initiation in Highly Irradiated Stainless Steels under Uniaxial Constant Load Conditions V. Chandravanshi , R. Sarkar , P. Ghosal et al. , Effect of minor additions of boron on microstructure and mechanical properties of as-cast near α titanium alloy . Metall. Mater. Trans. A 41 , 936 - 46 ( 2010 ). doi: 10.1007/s11661-009-0155-0 http://doi.org/10.1007/s11661-009-0155-0 . Effect of minor additions of boron on microstructure and mechanical properties of as-cast near α titanium alloy D. Dunning , W. Anderson , P. Mertens , Boron Containing Control Materials . Nucl. Sci. Eng 4 , 402 - 14 ( 1958 ). doi: 10.13182/NSE58-A25537 http://doi.org/10.13182/NSE58-A25537 . Boron Containing Control Materials A. Bhattacharya , C.M. Parish , T. Koyanagi et al. , Nano-scale microstructure damage by neutron irradiations in a novel Boron-11 enriched TiB2 ultra-high temperature ceramic . Acta Mater 165 , 26 - 39 ( 2019 ). doi: 10.1016/j.actamat.2018.11.030 http://doi.org/10.1016/j.actamat.2018.11.030 . Nano-scale microstructure damage by neutron irradiations in a novel Boron-11 enriched TiB2 ultra-high temperature ceramic Y. Sano , M. Takeuchi , Y. Nakajima et al. , Effect of metal ions in a heated nitric acid solution on the corrosion behavior of a titanium–5% tantalum alloy in the hot nitric acid condensate . J. Nucl. Mater 432 , 475 - 81 ( 2013 ). doi: 10.1016/j.jnucmat.2012.08.009 http://doi.org/10.1016/j.jnucmat.2012.08.009 . Effect of metal ions in a heated nitric acid solution on the corrosion behavior of a titanium–5% tantalum alloy in the hot nitric acid condensate N. Mattern , J. Han , O. Fabrichnaya et al. , Experimental and thermodynamic assessment of the Gd–Ti system . Calphad 42 , 19 - 26 ( 2013 ). doi: 10.1016/j.calphad.2013.06.007 http://doi.org/10.1016/j.calphad.2013.06.007 . Experimental and thermodynamic assessment of the Gd–Ti system N. Mattern , J. Han , M. Zinkevich et al. , Experimental and thermodynamic assessment of the Gd–Zr system . Calphad 39 , 27 - 32 ( 2012 ). doi: 10.1016/j.calphad.2012.08.001 http://doi.org/10.1016/j.calphad.2012.08.001 . Experimental and thermodynamic assessment of the Gd–Zr system H. Okamoto , Supplemental literature review of binary phase diagrams: Au-Ce, B-Pr, Bi-Gd, Bi-Ho, Cd-Sr, Ga-Ti, Gd-Pb, Gd-Ti, Mg-Mn, Mn-Nd, Nd-Ni, and Ni-Ti . J. Phase Equilib. Diffusion 36 , 390 - 401 ( 2015 ). doi: 10.1007/s11669-015-0389-z http://doi.org/10.1007/s11669-015-0389-z . Supplemental literature review of binary phase diagrams: Au-Ce, B-Pr, Bi-Gd, Bi-Ho, Cd-Sr, Ga-Ti, Gd-Pb, Gd-Ti, Mg-Mn, Mn-Nd, Nd-Ni, and Ni-Ti V. Mallipudi , S. Valance , J. Bertsch , Meso-scale analysis of the creep behavior of hydrogenated Zircaloy-4 . Mech. Mater 51 , 15 - 28 ( 2012 ). doi: 10.1016/j.mechmat.2012.03.003 http://doi.org/10.1016/j.mechmat.2012.03.003 . Meso-scale analysis of the creep behavior of hydrogenated Zircaloy-4 M. Rafique , S. Chae , Y.-S. Kim , Surface, structural and tensile properties of proton beam irradiated zirconium . Nucl. Ins Methods. Physics. Res B 368 , 120 - 8 ( 2016 ). doi: 10.1016/j.nimb.2015.12.001 http://doi.org/10.1016/j.nimb.2015.12.001 . Surface, structural and tensile properties of proton beam irradiated zirconium J. Dupont , C. Robino , T. Anderson , Influence of Gd and B on solidification behaviour and weldability of Ni–Cr–Mo alloy . Sci. Technol. Weld. Joi 13 , 550 - 65 ( 2008 ). doi: 10.1179/174329308X345083 http://doi.org/10.1179/174329308X345083 . Influence of Gd and B on solidification behaviour and weldability of Ni–Cr–Mo alloy R. Mizia , J. Dupont , C. V. Robino et al. , Corrosion Performance of a Gadolinium Containing Stainless Steel . J. Hepa 46 , S100 ( 2007 ). doi: 10.14359/7626 http://doi.org/10.14359/7626 . Corrosion Performance of a Gadolinium Containing Stainless Steel R. E. Mizia , T. E. Lister , P. J. Pinhero et al. , Development and testing of an advanced neutron-absorbing gadolinium alloy for spent nuclear fuel storage . Nucl. Tech 155 , 133 - 48 ( 2006 ). Development and testing of an advanced neutron-absorbing gadolinium alloy for spent nuclear fuel storage P. Fauvet , Corrosion issues in nuclear fuel reprocessing plants . Nucl. Corros. Sci. Eng 29 , 679 - 728 ( 2012 ). doi: 10.1533/9780857095343.5.679 http://doi.org/10.1533/9780857095343.5.679 . Corrosion issues in nuclear fuel reprocessing plants E. Patrito , R. Torresi , E. Leiva et al. , Potentiodynamic and AC impedance investigation of anodic zirconium oxide films . J. Electrochem. Soc 137 , 524 ( 1990 ). doi: 10.1149/1.2086492 http://doi.org/10.1149/1.2086492 . Potentiodynamic and AC impedance investigation of anodic zirconium oxide films C. Xia , Z. Zhang , Z. Feng et al. , Effect of zirconium content on the microstructure and corrosion behavior of Ti-6Al-4V-xZr alloys . Corros. Sci 112 , 687 - 95 ( 2016 ). doi: 10.1016/j.corsci.2016.09.012 http://doi.org/10.1016/j.corsci.2016.09.012 . Effect of zirconium content on the microstructure and corrosion behavior of Ti-6Al-4V-xZr alloys D. R. Knittel , A. Bronson , Pitting corrosion on zirconium—a review . Corros 40 , 9 - 14 ( 1984 ). doi: 10.5006/1.3579296 http://doi.org/10.5006/1.3579296 . Pitting corrosion on zirconium—a review P. Pedeferri , Corrosion Science and Engineering || Pitting Corrosion . Eng. Mater 11 , 207 - 30 ( 2018 ). doi: 10.1007/978-3-319-97625-9_11 http://doi.org/10.1007/978-3-319-97625-9_11 . Corrosion Science and Engineering || Pitting Corrosion F. Robinson , Pitting corrosion: cause, effect, detection and prevention . Anti-Corros. Method. Mater , ( 1960 ). doi: 10.1108/eb019748 http://doi.org/10.1108/eb019748 . Pitting corrosion: cause, effect, detection and prevention X. Guo , Y. Sun , K. Cui . Darkening of zirconia: a problem arising from oxygen sensors in practice . Sensor. Actuat. B-Chem 31 , 139 - 45 ( 1996 ). doi: 10.1016/0925-4005(96)80058-X http://doi.org/10.1016/0925-4005(96)80058-X . Darkening of zirconia: a problem arising from oxygen sensors in practice C. Morant , J. Sanz , L. Galan et al. , An XPS study of the interaction of oxygen with zirconium . Surf. Sci 218 , 331 - 45 ( 1989 ). doi: 10.1016/0039-6028(89)90156-8 http://doi.org/10.1016/0039-6028(89)90156-8 . An XPS study of the interaction of oxygen with zirconium D. Raiser , J. Deville. Study of XPS photoemission of some gadolinium compounds . J. Electron. Spectrosc. Relat. Phenom 57 , 91 - 7 ( 1991 ). doi: 10.1016/0368-2048(91)85016-M http://doi.org/10.1016/0368-2048(91)85016-M . J. Deville. Study of XPS photoemission of some gadolinium compounds D. Sarma , C. Rao , XPES studies of oxides of second-and third-row transition metals including rare earths . J. Electron. Spectrosc. Relat. Phenom 20 , 25 - 45 ( 1980 ). doi: 10.1016/0368-2048(80)85003-1 http://doi.org/10.1016/0368-2048(80)85003-1 . XPES studies of oxides of second-and third-row transition metals including rare earths I. Milošev , T. Kosec , H. Strehblow , XPS and EIS study of the passive film formed on orthopaedic Ti–6Al–7Nb alloy in Hank's physiological solution . Electrochim. Acta 53 , 3547 - 58 ( 2008 ). doi: 10.1016/j.electacta.2007.12.041 http://doi.org/10.1016/j.electacta.2007.12.041 . XPS and EIS study of the passive film formed on orthopaedic Ti–6Al–7Nb alloy in Hank's physiological solution M. Fan , X. Ren , F. Zhou et al. , Research and application of nuclear radiation protection materials . IOP Conf. Ser. Earth Environ . Sci, Kamakura, Japan , August 21-22, 2021 . 012093 . doi: 10.1088/1755-1315/859/1/012093 http://doi.org/10.1088/1755-1315/859/1/012093 . Research and application of nuclear radiation protection materials H. Jing , L. Geng , S. Qiu et al. , Research progress of rare earth composite shielding materials . J. Rare. Earth 41 , 32 - 41 ( 2023 ). doi: 10.1016/j.jre.2022.06.004 http://doi.org/10.1016/j.jre.2022.06.004 . Research progress of rare earth composite shielding materials All authors contributed to the study conception and design. Material preparation, data collection and analysis were performed by Cheng-Jie Du, Xiao-Gang Hu, Ping-Yi Guo, Xiao-Long Pan, and Jin-Ping Wu. The first draft of the manuscript was written by Xiao-Gang Hu and all authors commented on previous versions of the manuscript. All authors read and approved the final manuscript. 10.1007/s41365-023-01193-4 This work was supported by the National Natural Science Foundation of China (Nos.52201021 and 52101099), Key Research and Development Program of Shaanxi (2021GY-249, 2021GY-233), Natural Science Basic Research Program of Shaanxi (No. 2020JC-50), and Shaanxi Provincial Natural Science Youth Foundation (2022JQ-410). 2022-10-06 2023-01-09 2023-01-28 2023-03-21 2023-04-20 © China Science Publishing & Media Ltd. (Science Press), Shanghai Institute of Applied Physics, the Chinese Academy of Sciences, ChineseNuclear Society and Springer Nature Singapore Pte Ltd. 2023 This is a post-peer-review, pre-copyedit version of an article published in Nuclear Science and Techniques. The final authenticated version is available online at: http://dx.doi.org/10.1007/s41365-023-01193-4 http://dx.doi.org/10.1007/s41365-023-01193-4 . Springer Nature or its licensor (e.g. a society or other partner) holds exclusive rights to this article under a publishing agreement with the author(s) or other rightsholder(s); author self-archiving of the accepted manuscript version of this article is solely governed by the terms of such publishing agreement and applicable law. 2023 The data that support the findings of this study are openly available in Science Data Bank at https://www.doi.org/10.57760/sciencedb.07295 and http://resolve.pid21.cn/31253.11.sciencedb.07295. Data availability Nuclear Science and Techniques 3 34