High-level Nuclear Waste Research Articles

Toward a Fully Integrated Multiphysics Simulation Framework for Fusion Blanket Design

Fusion is an attractive clean-energy solution, thanks to its various advantages, such as reduced radioactivity, little high-level nuclear waste, ample fuel supplies, and increased safety. However, the harsh operating environment introduced by a complex fusion plasma system makes design and integration of fusion blankets incredibly challenging and time-consuming. This work focuses on developing a fully integrated multiphysics simulation framework based on an advanced open-source platform—the Multiphysics Object-Oriented Simulation Environment (MOOSE)—to alleviate the difficulties in fusion blanket design and integration. MOOSE is a massively parallel finite element/volume multiphysics simulation platform that has been widely adopted within the nuclear fission community. Even though fission and fusion are fundamentally different, they involve similar multiphysics phenomena. A fully integrated open-source multiphysics simulation framework tailored for the fusion blanket design will be implemented by leveraging the well-established multiphysics capabilities in MOOSE. Once successfully developed, this fully integrated framework will rapidly evaluate a blanket design concept and offer insights for subsequent iterations. As the first step, we will mainly aim to integrate neutronics analysis, system thermal hydraulics simulation, and full 3-D heat transfer calculations. The efficacy of the integrated framework will be verified using an innovative solid ceramic blanket design. While the project’s final goal is to enable a fully integrated multiphysics simulation platform for various fusion blanket concepts, this work, as a preliminary step, will mainly focus on a solid ceramic breeder helium-cooled blanket.

Impacts of substitution of Fe2O3 for SiO2 on structure and properties of borosilicate glasses containing MoO3

Borosilicate glass is an effective matrix for vitrification of high-level nuclear waste. However, the solubility of molybdenum is lower. In this paper, the impacts of Fe2O3 replacing SiO2 on the phase, glass structure, and chemical durability of a simplified calcium borosilicate glass containing molybdenum were investigated. According to XRD, Raman, and FTIR spectra, replacing SiO2 with Fe2O3 makes the glass structure polymerize first and then depolymerize. Furthermore, when Fe2O3 replaces SiO2 up to 5 mol%, the samples show a fully amorphous character, and the glasses show a more disordered structure. The solubility of Mo in the simplified calcium borosilicate glass can be enhanced by extending the depolymerization regions and introducing Fe3+ to compensate for the charge of [MoO4]2−. Moreover, the substitution of Fe2O3 can improve the chemical durability of glass waste form. As a result, the borosilicate glass containing 6 mol% Fe2O3 is a potential matrix to immobilize Mo-containing HLW.

Enhanced Crevice Corrosion of Stainless Steel 316 By Degradation of Cr-Containing Hollandite Crevice Former

The disposal of high-level nuclear waste (HLW) is an important societal problem. It is also an extremely challenging corrosion research topic due to the highly complicated waste chemistry, long lasting radiation effect, and inevitable exposure to aqueous environments for hundreds of thousands of years. To safely accommodate some critical radionuclides such as volatile 129I and strongly heat generating 137Cs, numerous crystalline ceramic waste forms have been created and studied. Among the promising nuclear waste forms, hollandite was developed to isolate Cs waste and the corresponding decay product Ba due to the open tunnel structure capable of accommodating mono- or divalent cations. In this study, the synergistic corrosion interactions between a Cr-containing hollandite (Ba1.15Cr2.3Ti5.7O16) and stainless steel (SS) 316 is explored. This is relevant to the permanent disposal of HLW involving the encasement of glass or crystalline ceramic waste forms containing immobilized radionuclides in a metallic canister made from corrosion resistant alloys such as SS. The experiments performed in this study simulate the potential corrosion interactions occurring at the interface of the SS and ceramic. After corroding the SS316 and Cr-containing hollandite in proximity in 0.6 M NaCl at 90 oC for 28 days, severe crevice corrosion was identified on the surface of the SS, as evidenced by the presence of large pits and crevice damage with diameters of hundreds of microns. Similarly, localized damage was also found at matching sites on the Cr-hollandite surface, indicating interactions between the two materials. The synergistic corrosion interaction was likely due to the continuous release of Cr3+ cations from both materials, including the passive dissolution of the SS and the heterogenous degradation of the Cr-hollandite. The extra Cr3+ cations originated from the hollandite reduces the incubation time required to reach the critical crevice solution condition, thereby accelerating the breakdown of the passive film of SS and the subsequent onset of active dissolution. An adapted crevice corrosion model is applied to quantitively explain the accelerated corrosion of SS observed in this study.

Energy, Economical, Environmental and Medical Applications of Cold Nuclear Fusion of Hydrogen with Powder and Liquid Forms of Metals

Aim of this paper is to highlight the four major applications of Cold Nuclear fusion (CNF). Considering Google publications, NASA publications, European funding amounts and other important experimental results, we have developed a mechanism for understanding the mystery of energy liberation in CNF associated with fusion of hydrogen with metals. Apart from clean thermal energy liberation, we have identified three important applications of CNF. They are: Preparing Iridium-, Platinum-and Gold-like costly elements from Tungsten-like elements, converting high level nuclear radioactive waste into stable atomic nuclides and preparing radioactive isotopes for medical applications. It may be noted that, we have developed a common, simple and workable theoretical procedure for understanding the above applications and it needs funding for conducting experimental studies. We emphasize the point that, our proposed scientific concept of CNF is practical compared to other theoretical models of cold nuclear fusion.

A systematic review of bentonite/concrete interaction system in HLW disposal repositories: Theoretical, experimental and geochemical modelling analysis

The buffer/backfill material of bentonite acts the most important role for the safety of high-level nuclear waste repository. After the closure of repository, the concrete as another barrier in geological repositories will degrade and release alkaline components into the site water, which has significant impact on the compositions of bentonite and threatens the long-term safety of repository. Many studies have been carried out to detect the mineralogy alteration of bentonite during alkaline disturbing including laboratory tests, in situ experiments and geochemical modelling. It was found that montmorillonite dissolves and forms low-swelling minerals such as mixed layer clays, non-swelling zeolite and calcium silicate hydrate minerals, etc. The researches that have been done are extensive and disorganized, however, the results are mostly qualitative and far from the engineering requirements.

Sorption Studies of Cesium on Kunipia-F Bentonite and Silicon Dioxide

Adsorption is one of the effective methods used for the removal of contaminants from radioactive wastewater. In this research, removal efficiency of Cs from the prepared low-concentration Cs solution (100 μg/L) was studied using two different adsorbents namely high purity Kunipia-F bentonite and Silicon dioxide (SiO2). After performing the adsorption tests, the Cs ions in liquid and solid phases were analyzed using the inductively coupled plasma mass spectrometry (ICP-MS) technique. The adsorption capacities of both materials were then compared. It was found that the Cs adsorption capacity of bentonite was higher than that of SiO2. The distribution coefficient of Cs in the bentonite was 1.0 L-sol/g-solid, which was 100 times greater than that of SiO2. The highest sorption of Cs on bentonite could be attributed to its smaller particle size (0.362 μm), compared to the SiO2 particle size (0.625 μm). Bentonite is natural clay and is generally used as a buffer material in geological disposal systems for high-level nuclear waste. The results of this research indicate that bentonite could be an effective absorbent for decontaminating low concentration Cs in radioactive wastewater.

Understanding the Mechanism of Cs Accumulation on Stainless Steel Suspended in Nuclear High-Level Liquid Waste.

In a nuclear facility, the surface of stainless steels (SS) was found to be contaminated during the processing of radioactive liquid waste. Their safe and secure disposal is highly challenging to the nuclear industry. If the fundamental property of steel corrosion could be evaluated, successful decontamination and effective decommissioning strategies could be planned. Although individual radionuclide contamination behavior on SS metal was studied, till date, SS contamination behavior under the exposure of high-level liquid waste (HLLW) was unexplored. In view of this, investigations were carried out to understand the nature of contamination in SS 304L alloy under the exposure of simulated HLLW (SHLLW). For understanding of radionuclide adsorption behavior on structural materials, scanning electron microscopy/energy dispersive X-ray spectroscopy and X-ray photoelectron spectroscopy have been utilized for SS 304L. The solutions were analyzed using inductively coupled plasma optical emission spectroscopy to calculate the changes in the elemental composition of the solution and corrosion behavior of SS. The passivation of SS coupons was observed in the presence of HNO3 due to enrichment of Cr at the surface. The deposition of Cs and Mo was noticed, while SS coupons were exposed to SHLLW. At 3 M HNO3 and room temperature, the SS surface is mildly passivated by Cr enrichment by formation of a Cr oxide layer on the SS surface. However, the passive layer was not thick enough to attenuate the signal from the substrate below the passivated layer. Hence, Fe0 and Cr0 were also found along with Cr3+ and Fe3+ (in small quantity). When temperature was elevated to 70 °C, the SS surface was completely covered with the Cr oxide layer, and hence no Cr0 signal was observed. The small signal of Fe0 indicated that the signal from the substrate surface is present below the passive layer. During the passivation process, Cr diffused toward the passive layer, thereby producing a Cr-depleted layer below the passive layer (Cr0 signal was not seen). In the case of SHLLW at 70 °C, the surface was fully covered by Cr3+, Mo6+, and Cs+. Fe and Ni were not observed at all. This finding will help to design an effective corrosion inhibitor and suitable decontamination agent. In addition to that, this information was found to be useful in designing high-performance novel and modern age reactor materials with improved characteristics.

Comparative study on the reductive immobilization of Se(IV) by Beishan granite and Tamusu claystone

Understanding the retardation of 79Se by the host rock is essential for the safety assessment of the high-level radioactive waste (HLRW) repository and thus the final site-selection process. In this work, we performed a comparative study on the interaction of Se(IV) with Beishan granite and Tamusu claystone, and multiple characterization techniques were used to identify the speciation of Se and to unravel the underlying mechanism. Results showed that Se(IV) can be reduced to Se(0) by the Beishan granite from borehole 16 that contains Fe(II)-bearing fluorannite, and the removal efficiency of Se(IV) is more favorable at acidic condition. We proposed that release of the structure Fe2+ in granite, followed by surface adsorption and reaction could be responsible for the Se(IV) reduction by this granite. In addition to Se(0) product, FeSe2 was also observed for Se(IV) reacted with the granite from borehole 28. This is the first time observed the formation of FeSe2 on natural Fe(II)-containing minerals, which is mainly ascribed to the simultaneous occurrence of aqueous sulfide dissolved from the ground granite powder. The reactive sulfide might be generated from the breakage of Fe(S)–S bonds of iron sulfides (e.g., pyrite) contained in the granite during the anoxic grinding process, and it could mainly account for the Se(IV) reduction by the granite from borehole 28. Due to the presence of pyrite, the Tamusu claystone from borehole Tzk2 was effective to remove aqueous Se(IV) via the formation of Se(0). Conversely, the Se(IV) removal efficiency was much lower for the claystone from borehole Tzk1 that did not contain pyrite, although there was a considerable amount of Fe(II)-bearing fluorannite. Hence, the Fe2+ contained in Tamusu claystone was less active for Se(IV) reduction in comparison with that in Beishan granite from borehole 16. The findings of this study provide important data for the safety assessment of China's future HLRW repository, but also improve our understanding of the geochemical behavior of selenium in the presence of Fe(II)- and sulfide-bearing minerals in the relevant anoxic environments.

Study of Radium Behavior in Contact With Calcium-Silicate-Hydrates

Abstract Radium-226 is an important radionuclide with respect to the long-term safety of repositories for low and intermediate-level radioactive wastes, as well as deep geological repositories for high-level radioactive waste and spent nuclear fuel. To evaluate its mobility in the engineered barrier system of a repository, the sorption of radium on calcium-silicate-hydrates (CSH), which form the major hydration phases in cementitious materials, was studied. Radium sorption to CSH was found to be very fast, leading to steady-state between solid and liquid phase after less than four days. The dependency of the equilibrium distribution ratios on radium concentration, the calcium to silicon ratio in CSH, liquid to solid ratio, and temperature was investigated, and estimates of apparent activation energy, reaction enthalpy, entropy, and Gibbs energy of the sorption process were derived. Radium sorption on CSH can be described by linear isotherms with Rd values mostly in the order of 104 L/kg. Radium sorption was found to be an exothermic and spontaneous reaction probably governed by chemical reaction rather than diffusion. As expected, the presence of ethylene diamine tetraacetic acid (EDTA) at low concentrations led only to a small decrease in radium sorption, due to the strong competition of dissolved calcium for EDTA complexation. A comparison of the sorption behavior of various alkaline earth elements used as chemical analogs for radium confirmed the significant difference in the Rd values with sorption on CSH decreasing in the order Ra > Ba > Sr.

Surface interfacial analysis of simulant high level nuclear waste glass dissolved in synthetic cement solutions

The corrosion mechanisms and kinetics of a Mg-rich alkali aluminoborosilicate glass simulating UK high-level waste (CaZn28) were investigated upon dissolution in synthetic cement solutions. Dissolution varied as a function the different pH and alkali/alkaline earth content of each cement solution. High resolution microscopy and spectroscopy techniques ascertained the nature of the interface between the glass and the cement solutions. TEM-EDS revealed alkali- and alkaline earth-rich silica gels, into which K, Ca and Mg were incorporated. TEM-SAED, combined with synchrotron micro-focus XRD, identified the ubiquitous precipitation of the Mg-aluminate layered double hydroxide phase, meixnerite (Mg6Al2(OH)18·4H2O), in addition to goethite (FeOOH) and crystalline silica. The C-S-H phase, tobermorite (Ca5Si6O16(OH)2·4H2O), was identified in the most Ca-rich solution only. These data give insight to the role of alkali/alkaline earth-rich solutions in the dissolution or radioactive waste glasses, of importance to the final disposition in a geological disposal facility.

Effect of thermal ageing on physical properties of MX80 bentonite under high-temperature conditions

In a deep geological repository (DGR) system, the buffer layer is indispensable to ensuring the safe disposal of high-level radioactive nuclear waste (HLW). Because the heat generated by the decay of the spent nuclear fuel in a canister is released to the surrounding buffer layers, the bentonite buffer material experiences long-term high-temperature conditions. Therefore, the variations in physical properties of bentonite buffer material under high-temperature conditions are one of the important parameters in the DGRs for HLW. A series of tests on the specific gravity, specific surface area (SSA), Atterberg's limits, and free swelling ratio of MX80 bentonite after heating for different times at a high temperature of 200 °C were conducted, to investigate the influence of thermal ageing time on its physical properties. Then, the microscopic investigations, including X-ray diffraction (XRD), thermogravimetric analysis (TGA), and scanning electron microscopy (SEM), were conducted to explain the aforementioned variations from the microscopic point of view. Based on the crystal layer structure characteristics of montmorillonite, the quantitative relationships between the physical properties of bentonite and its mineral composition or bound water content were established to further explain the mechanism by which thermal ageing affects the physical properties of MX80 bentonite. The results indicate that with increasing heating time, the specific gravity, SSA, liquid limit, plastic limit, plastic index and free swelling ratio decrease sharply by 2.5%, 4.5%, 3.1%, 5.5%, 2.8% and 30.2%, respectively, within 15–30 days of heating. After 30 days of heating, their variations are negligible. Under high-temperature conditions, the transformation of mineral composition, desorption of bound water, and changes in the micro-morphology are the fundamental reasons for the variations in the physical properties of bentonite, and they influence and interact with each other.

Insight on the Void Ratio-Suction Relationship of Compacted Bentonite during Hydration.

Investigation on swelling characteristics of buffer/backfill materials during hydration is an important issue in the design of artificial barriers in high-level radioactive waste (HLW) disposal repositories. In this work, for clarifying the characteristic of void ratio–suction relationship for compacted bentonite on hydration path, suction-controlled swelling deformation tests under constant vertical stresses 0.001~40 MPa were carried out on compacted bentonite specimens. Four different types of void ratio-suction curves indicated that swelling-collapse behavior under hydration depends on suction and over-consolidation ratio (OCR), based on which the swelling index was defined. Then, equations were proposed for describing the swelling-collapse characteristic of void ratio–suction curves. Simulation results of suction-controlled swelling deformation tests show that the different types of the hydration deformation curves could be well described by the proposed equations. Obviously, the proposed equations could be used for description and prediction of swelling characteristics of compacted bentonite during hydration, which is also of great importance for the safety assessment of the HLW repositories.

Response of structure and mechanical properties of high entropy pyrochlore to heavy ion irradiation

Material with superior damage tolerance, chemical durability, and structure stability is of increasing interest in high-level radioactive waste management and structural components for advanced nuclear systems. In this paper, high-entropy (La0.2Ce0.2Nd0.2Sm0.2Gd0.2)2Zr2O7 with pyrochlore-type structure was synthesized through conventional solid-state method. The as-synthesized high-entropy oxide maintained crystalline after being irradiated by using Au3+ with 9.0 MeV energy at the fluence of 4.5 × 1015 ions·cm-2, indicating its high tolerance to heavy-ion irradiation. The irradiation-induced order-disorder transition from pyrochlore structure to defective fluorite structure occurred in high-entropy (La0.2Ce0.2Nd0.2Sm0.2Gd0.2)2Zr2O7. After irradiation, no irradiation-induced segregation was observed at grain boundary. Moreover, the mechanical properties of high-entropy pyrochlore were improved. The heavy-ion irradiation resistance mechanisms of high-entropy pyrochlore were discussed in detail. Our work identified high-entropy (La0.2Ce0.2Nd0.2Sm0.2Gd0.2)2Zr2O7 can be a promising candidate for immobilization of high-level radioactive waste as well as advanced nuclear reactor system from the perspective of irradiation resistance.

Characterization and modeling of chemical reactions taking place during the vitrification of high level nuclear waste

Several chemical reactions take place during the vitrification of high-level nuclear waste. The integration of these chemical aspects in process numerical simulation tools is one of the key points in the approach of improving the prediction capabilities of numerical tools today. In this study, based on simultaneous differential scanning calorimetry-thermogravimetry (DSC-TGA) and the run/rerun method, we propose a modeling of the kinetics of chemical reactions taking place when nuclear glass precursors are subjected to constant heating. For the mathematical modeling, a weighted sum of nth order reactions was used to describe the overall mechanism. A hybrid approach coupling the Kissinger and least squares methods was performed to identify the apparent kinetic parameters. Along with the thermal characterization, we also provide some information about the nature of the main reactions through microstructural evolution and evolved gas analyses.

Uncertainty quantification of radionuclide migration in fractured granite

Deep geological disposal is a widely accepted approach for safe management and long-term disposal of high-level radioactive waste (HLW). However, high uncertainty associated with subsurface properties of fractured rocks is a significant obstacle to practical safety assessment of HLW disposal. In this study, we develop an integrated statistical framework for uncertainty quantification of radionuclide migration related to the geological disposal of HLW. We employ a response surface methodology integrated with Monte Carlo simulations of radionuclide migration in fractured granite to perform global sensitivity and statistical analysis by coupling the uncertainty quantification tool, PSUADE, and radionuclide migration simulator, FRACPIPE. FRACPIPE is a semi-analytical simulator that models solute transport in fractures and matrix slabs with an arbitrary-length decay chain and an arbitrary time-varying influent concentration history by considering a variety of transport mechanisms. The statistical risk metrics include nuclide breakthrough (BT) time, total release dose, single nuclide release dose, and single nuclide flux. The global sensitivity analysis identifies fracture aperture, matrix diffusion coefficient, hydraulic gradient, and dispersivity as the most sensitive parameters. Considering the uncertainty ranges for independent variables (e.g., dispersivity, hydraulic gradient, fracture spacing, fracture aperture, sorption distribution coefficient, matrix diffusion, and matrix porosity), we apply post-processing results of Monte Carlo simulations to conduct statistical analysis of the risk metrics. Under this studied condition, the average BT time for radionuclides at 1000 m is between 85 and 331 thousand years. This study provides valuable insights into the impact of input parameter (independent variables) uncertainty and sensitivity on radionuclide migration behavior in an HLW repository.

Long-term interactive corrosion between International Simple Glass and stainless steel

The interactive corrosion between stainless steel 316 (SS) and International Simple Glass (ISG) has been studied for 12 months. The samples were placed in intimate contact to create crevice corrosion conditions on SS in a 0.6 M NaCl solution with and without 142 ppm dissolved silica species enriched in 29Si. Corrosion on SS was observed with large pits located close to the crevice mouth at the end of 1 year in 0.6 M NaCl solution. The SS corrosion was also associated with the dissolution of nearby glass in a highly localized manner. When the leaching solution was saturated with silica species, the corrosion rate of SS was suppressed. However, the presence of SS enhances ISG dissolution in both conditions. The implications of the observed phenomena on the long-term storage of nuclear waste are discussed.

Law of Nuclide Migration in Clayey Rocks considering Diffusion and Fluid Transport

A core concern in the research on deep geological disposal of high-level radioactive waste is the migration of radionuclides in geological bodies. Most studies on radionuclide migration consider the role of only the rock fissures without incorporating the influence of the rock matrix. In this paper, the rock mass for geological disposal of high-level radioactive waste is regarded as a fissure-pore medium. Considering the influences of radionuclide diffusion and fluid transport on radionuclide migration in the process of disposal, the governing equation of radionuclide migration and evolution in the pore-fissure medium is established. The numerical scheme of the governing equation is given based on the mixed finite volume method (FVM), using our program solution module written in C++. On this basis, the numerical test model with fissures was developed, which analyzed the radionuclide migration law in clayey rocks under various fissure and rock matrix diffusion coefficients and hydraulic conductivities. The simulation results are compared with finite element method results, revealing the superiority of the mixed FVM method in solving problems of radionuclide migration in discontinuous geological bodies containing hiatuses, mutations, and fissures. The study provides a theoretical basis for evaluating the safety, feasibility, and suitability of geological disposal repositories for high-level radioactive waste in terms of radionuclide migration.

Re-passivation rate and conduction mechanism of surface film on copper in nitrite solutions

In this work, an effort was put to verify the usefulness of the potential step re-passivation technique in studying the stress corrosion cracking (SCC) susceptibility of copper in environments simulating the final repository of high-level nuclear waste. Validation experiments in NaNO2 solutions (0.005, 0.05 and 1.0 M) showed that the technique is applicable with certain boundary conditions. In addition, conduction mechanism of passive films on Cu in nitrites is studied with impedance spectroscopy and quantitatively modeled by the Mixed-Conduction Model for oxide films. As a result, possible correlations between electrochemically derived parameters and SCC susceptibility are found and discussed.

Evaluation on the buffer temperature by thermal conductivity of gap-filling material in a high-level radioactive waste repository

As high-level radioactive waste (HLW) generated from nuclear power plants is harmful to the human body, it must be safely disposed of by an engineered barrier system consisting of disposal canisters and buffer and backfill materials. A gap exists between the canister and buffer material in a HLW repository and between the buffer material and natural rock—this gap may reduce the water-blocking ability and heat transfer efficiency of the engineered barrier materials. Herein, the basic characteristics and thermal properties of granular bentonite, a candidate gap-filling material, were investigated, and their effects on the temperature change of the buffer material were analyzed numerically. Heat transfer by air conduction and convection in the gap were considered simultaneously. Moreover, by applying the Korean reference disposal system, changes in the properties of the buffer material were derived, and the basic design of the engineered barrier system was presented according to the gap filling material (GFM). The findings showed that a GFM with high initial thermal conductivity must be filled in the space between the buffer material and rock. Moreover, the target dry density of the buffer material varied according to the initial wet density, specific gravity, and water content values of the GFM.

Adsorption characteristics of strontium by bentonite colloids acting on claystone of candidate high-level radioactive waste geological disposal sites

The bentonite colloid produced in the deep geological repository of high-level radioactive waste can directly affect the migration of radionuclide strontium when it acts on claystone. The adsorption characteristics of strontium were investigated on claystone with the presence or absence of bentonite colloids from the Suhongtu area of China. The effects of different influencing factors, such as pH and solid content, on the adsorption process were investigated by batch adsorption experiments, and spectroscopic techniques were used to characterize the samples before and after adsorption of strontium. The results show that the presence of bentonite colloids can promote the adsorption of strontium on claystone under alkaline conditions. and the general order kinetic model provided the best fit to the experimental data. Strontium is adsorbed on the surface of claystone and bentonite colloid by ion exchange and surface complexation. Most of the Sr2+ formed SrCO3 with CO32− after ion exchange with Ca2+ and Mg2+ in plagioclase and dolomite, and a small amount of Sr2+ was adsorbed by complexation with –OH, Al–O and Si–O. These results provide a scientific basis for predicting the migration of strontium in subsurface porous media and the siting of high-level radioactive waste repositories.