Reduction of plutonium(VI) in brine under subsurface conditions

Plutonium Oxidation States in the Waste Isolation Pilot Plant Repository

The Waste Isolation Pilot Plant (WIPP), a deep geologic repository located 660 meters underground in bedded salt, is designed to isolate U.S. defense-related transuranic waste from the accessible environment. Plutonium isotopes are the most important radionuclides in WIPP waste. Plutonium solubility in WIPP brines (ionic strengths from 5.3 to 7.4) is strongly dependent on its oxidation state, with much lower solubilities associated with Pu(III) and Pu(IV) than with the higher Pu(V) and Pu(VI) oxidation states. The large quantity of metallic iron in WIPP waste and waste containers is expected to undergo anoxic corrosion, producing strongly reducing conditions and high hydrogen gas pressures after repository closure and brine intrusion. Because reducing conditions will prevail in the WIPP repository, the most important long-term oxidation states will be Pu(III) and Pu(IV). We performed a literature review to evaluate the effects of WIPP chemical and physical processes (not colloidal) on plutonium oxidation states that included reactions with reducing agents such as iron solids and aqueous species and radiolysis of solids and aqueous species. The results of this review indicate that equilibrium between Pu(III) solids and aqueous species will control dissolved plutonium concentrations in WIPP brines. We also performed geochemical modeling calculations using the ThermoChimie database to support this assessment of plutonium oxidation states in the long-term WIPP repository. Control of plutonium solubilities by Pu(III) solid instead of Pu(IV) solid may lead to higher predicted plutonium concentrations in brines potentially released to the ground surface by an inadvertent drilling intrusion into the long-term WIPP repository. The results of this study demonstrate that Pu(III) solid solubilities provide a reasonable upper bound for dissolved plutonium concentrations in WIPP brines.

Plutonium environmental chemistry: mechanisms for the surface-mediated reduction of Pu(v/vi)

In recent decades, interest in plutonium mobility has increased significantly due to the need of the United States, as well as other nations, to deal with commercial spent nuclear fuel, nuclear weapons disarmament, and the remediation of locations contaminated by nuclear weapons testing and production. Although there is a global consensus that geologic disposal is the safest existing approach to dealing with spent nuclear fuel and high-level nuclear waste, only a few nations are moving towards implementing a geologic repository due to technical and political barriers. Understanding the factors that affect the mobility of plutonium in the subsurface environment is critical to support the development of such repositories. The importance of redox chemistry in determining plutonium mobility cannot be understated. While Pu(iv) is generally assumed to be immobile in the subsurface environment due to sorption or precipitation, Pu(v) tends to be mobile due to its relatively low effective charge and weak complex formation. This review highlights one particularly important aspect of plutonium behaviour at the mineral-water interface-the concept of surface-mediated reduction, which describes the reduction of plutonium on a mineral surface. It provides a conceptual model for and evidence supporting or refuting each proposed mechanism for surface-mediated reduction including (i) radiolysis at the mineral surface, (ii) electron transfer via ferrous iron or manganese in the mineral structure, (iii) electron shuttling due to the semiconducting properties of the mineral, (iv) disproportionation of Pu(v), (v) facilitation by proton exchange sites, (vi) stabilisation of Pu(iv) due to the increased concentration gradient within the electrical double layer, and (vii) a Nernstian favourability of Pu(iv) surface complexes and colloids. It also provides new perspectives on future research directions.

Exploring actinide materials through synchrotron radiation techniques

Synchrotron radiation (SR) based techniques have been utilized with increasing frequency in the past decade to explore the brilliant and challenging sciences of actinide-based materials. This trend is partially driven by the basic needs for multi-scale actinide speciation and bonding information and also the realistic needs for nuclear energy research. In this review, recent research progresses on actinide related materials by means of various SR techniques were selectively highlighted and summarized, with the emphasis on X-ray absorption spectroscopy, X-ray diffraction and scattering spectroscopy, which are powerful tools to characterize actinide materials. In addition, advanced SR techniques for exploring future advanced nuclear fuel cycles dealing with actinides are illustrated as well.

Thermodynamic description of Np(VI) solubility, hydrolysis, and redox behavior in dilute to concentrated alkaline NaCl solutions

The solubility of Np(VI) was investigated in carbonate-free NaCl solutions (0.1 M ≤ I ≤ 5.0 M) at T = 22 ± 2 °C to derive thermodynamic properties of aqueous species and solid compounds formed under alkaline conditions. The experimentally derived solubility curves can be divided into four main regions: (I) ~7 ≤ pHm ≤ ~9, showing a steep decrease in Np solubility with a slope (log [Np] vs. pHm) of –3 or –2 (depending on NaCl concentration); (II) ~9 ≤ pHm ≤ ~10.5, with a nearly pH-independent [Np]; (III) ~10.5 ≤ pHm ≤ ~13.5, showing an increase in the solubility with a well-defined slope of +1. A region (IV) with a slope ≥ +2 was only observed at I ≥ 1.0 M NaCl and pHm ≥ ~13.5. The solubility-controlling solid Np phases were characterized by X-ray diffraction (XRD), quantitative chemical analysis, thermogravimetric analysis and scanning electron microscopy-energy-dispersive spectrometry (SEM-EDS), confirming the presence of anhydrous Na2Np2O7(cr) in regions II and III. The same solid phase was identified in region I except for the system in 0.1 M NaCl, where a NpO2(OH)2·H2O(cr) phase predominates. XRD patterns of this solid phase show a very good agreement with that of metaschoepite (UO3·2H2O), highlighting the similarities between Np(VI) and U(VI) with respect to solid phase formation and structure. Based on the analysis of solubility data, solid phase characterization and chemical analogy with U(VI), the equilibrium reactions 0.5 Na2Np2O7(cr) + 1.5 H2O ⬄ Na+ + NpO2(OH)3– and 0.5 Na2Np2O7(cr) + 2.5 H2O ⬄ Na+ + NpO2(OH)42– + H+ were identified as controlling Np(VI) solubility in regions II and III, respectively. The predominance of NpO2+ in the aqueous phase of region I (quantified by UV–vis/NIR) indicates the reductive dissolution of Np(VI) [either as Na2Np2O7(cr) or NpO2(OH)2·H2O(cr)] to Np(V)aq. Oxidation to Np(VII) can explain the experimental observations in region IV, although it is not included in the chemical and thermodynamic models derived. The conditional equilibrium constants determined from the solubility experiments at different ionic strengths were evaluated with both the specific ion interaction theory (SIT) and Pitzer approaches. Thermodynamic data for aqueous Np(VI) species [NpO2(OH)3– and NpO2(OH)42–] and solid compounds [Na2Np2O7(cr) and NpO2(OH)2·H2O(cr)] that are relevant under alkaline conditions were derived. These data are not currently included in the Nuclear Energy Agency-Thermochemical Database (NEA-TDB) compilation.

A comparative study for radiological decontamination of laboratory fume hood materials

The efficacy for radiological decontamination of the laboratory standard fume hood as constructed of stainless steel, compared to that of powder-coated carbon steel is described. While the chemical inertness of powder-coated surfaces is good, faced with everyday abrasion, aggressive inorganic solutions and vapors, and penetrating organics commonly employed in government laboratory fume hoods, radiological decontamination of powder-coated steel surfaces was found to be similar to those made of stainless steel for easily solubilized or digestible radionuclides. Plutonium was difficult to remove from stainless steel and powder-coated surfaces, especially after prolonged contact times.