Momentum informed muon scattering tomography for monitoring spent nuclear fuels in dry storage cask

Development of an effective monitoring method for spent nuclear fuel (SNF) in a dry storage cask (DSC) is important to meet the increasing demand for dry storage investigations. The DSC investigation should provide information about the quantity of stored SNF, and quality assurance of materials should be possible without opening the cask. However, traditional nondestructive examination (NDE) methods such as x-rays are difficult to deploy for DSC investigation because a typical DSC is intentionally designed to shield against radiation. To address this challenge, cosmic ray muons (CRMs) are used as an alternative NDE radiation probe because they can easily penetrate an entire DSC system; however, a wide application of muons is often hindered due to the naturally low CRM flux (~104 muons/m2/min). This paper introduces a newly proposed imaging algorithm, momentum-informed muon scattering tomography (MMST), and presents how a limitation of the current muon scattering tomography technique has been addressed by measuring muon momentum. To demonstrate its functionality, a commercial DSC with 24 pressurized light water reactor fuel assemblies (FAs) and the MMST system were designed in GEANT4. Three noticeable improvements were observed for MMST system as a DSC investigation tool: (1) a signal stabilization, (2) an enhanced capability to differentiate various materials, and (3) statistically increased precision to identify and locate missing FAs. The results show that MMST improves the investigation accuracy from 79 to 98% when one FA is missing and 51% to 88% when one-half FA is missing. The advancement of the NDE technique using CRM for DSC verification is expected to resolve long-standing problems in increasing demand for DSC inspections and nuclear security.

Dataset of observables for UOX and MOX spent fuel extracted from Serpent2 fuel depletion calculations for PWRs

This database contains the isotopic mass density and the contribution to activity, decay heat, photon emission, spontaneous fission rate, (α,n) emission rates and radiotoxicity of 150 nuclides that are present in nuclear fuel irradiated in PWRs. These nuclides are of paramount importance for nuclear waste characterization and fuel cycle analysis. These values were obtained by depletion calculations based on a 3D pin-cell geometry model and performed with the Monte Carlo reactor physics burnup calculation code Serpent2, with state-of-the-art nuclear data libraries and relevant methods. The calculations cover a wide range of burnup levels for conventional PWRs and take into account both UOX and MOX fuel. A broad span for initial enrichment for UOX (from 1.5% to 6.0%), and for both the initial plutonium content (from 4.0% to 12.0% and the plutonium isotopic composition of MOX has been considered. This database has been made publicly available due to its relevance in the fields of waste and fuel characterization, nuclear safeguards and radiation protection, and it will allow other potential users to avoid the time-consuming calculations required to obtain the aforementioned data. Additionally, it constitutes an interesting dataset for model training in machine learning applications related to nuclear science and engineering.

Sequestration of cobalt and nickel by biofilm forming bacteria isolated from spent nuclear fuel pool water

In the current study, six bacterial types, isolated from spent nuclear fuel (SNF) pool facility, were investigated for their ability to sequester heavy metals (cobalt and nickel). Biofilm formation by the six bacterial isolates, viz., Bacillus subtilis, Staphylococcus species, Staphylococcus arlettae, Staphylococcus epidermidis, Staphylococcus auricularis, and Chryseobacterium gleum, were assayed, and they were found to have significant biofilm forming property. Their biofilms were characterised using confocal scanning laser microscopy, and their potential to accumulate Co²⁺ and Ni²⁺ from bulk solutions was analysed with respect to time. A comparative assessment of bioaccumulation capacity was done using biofilms, planktonic cells, and live vs dead cells. The strains accumulated Co²⁺ and Ni²⁺ in the range of 4 × 10^-4 to 1 × 10^-5 g/mg of cell biomass. It is interesting to note that dead biomass also showed significant removal of the two metal ions, suggesting an alternative process for metal removal. This study suggests that hostile environments can be a repertoire of putative bacterial species with potential heavy metals and other contaminants remediation properties.

Quantification of elements in spent nuclear fuel using intrinsic radioactivity for sample excitation and chemometric data processing

Quantitative analysis of spent nuclear fuel (SNF) is a very challenging task. High radioactivity, complex chemical composition and personnel safety requirements severely limit the number of analytical tools suitable for this problem. There is an urgent need for the methods that would provide for remote on-line quantification of elements in spent nuclear fuel and its reprocessing technological solutions. Here we propose a novel approach based on the registration of X-ray fluorescence radiation from SNF samples induced by fission products radioactivity. In this case the X-ray excitation conditions will obviously vary from sample to sample; moreover the resulting spectra will be a complex superposition of numerous signals from soft gamma emitters and X-ray fluorescence of various nature. These complex spectra can be effectively treated with chemometric data processing for quantification of particular elements. We have demonstrated the validity of this approach for direct analysis of U, Zr and Mo in SNF raffinate.

Quantification of uranium, plutonium, neodymium and gadolinium for the characterization of spent nuclear fuel using isotope dilution HPIC-SF-ICP-MS

A method was developed for the determination of the nuclide-specific concentrations of U, Pu, Nd and Gd in two types of spent nuclear fuel (UOx and Gd-enriched). High-performance ion chromatography (HPIC) was used to separate the target elements from one another while sector-field inductively coupled plasma-mass spectrometry (SF-ICP-MS) was used for their determination relying on isotope dilution for calibration. In order to obtain the best possible precision for these isotope ratios extracted from the transient HPIC-SF-ICP-MS signals, the SF-ICP-MS data acquisition parameters were optimized and the most suitable method for calculating the isotope ratios from the transient signals was identified. The point-by-point (PbP), linear regression slope (LRS) and peak area integration (PAI) approaches were compared in the latter context. It was found that data acquisition in the flat centre of the spectral flat top peak using a mass window of 25%, a dwell time of 10 ms and 20 samples per peak, while using PAI for isotope ratio calculation, gave the best precision on the isotope ratios extracted from the HPIC-SF-ICP-MS transient signals. These parameters were used in the determination of the nuclide-specific mass fractions of Pu, Nd and Gd in two types of spent nuclear fuel using isotope dilution HPIC-SF-ICP-MS. For U, which was present at a higher concentration, the element fraction was collected and analyzed off-line after dilution. For the other target elements, an online approach was used. An uncertainty budget estimation was made using the bottom-up approach for the resulting mass fractions, and the accuracy and precision obtained when using isotope dilution HPIC-SF-ICP-MS were compared with those obtained with the routinely used techniques, isotope dilution TIMS & alpha spectrometry (an ISO 17025 accredited method).

Pressurized water reactor spent nuclear fuel data library produced with the Serpent2 code

The paper describes a data library containing material composition of spent nuclear fuel. The data is extracted from burnup and depletion calculations with the Serpent2 code. The simulations were done with a PWR fuel pin cell geometry, for both initial UO₂ and MOX fuel load for a wide range of initial enrichments (IE) or initial plutonium content (IPC), discharge burnup (BU) and cooling time (CT).

The fuel library contains the atomic density of 279 nuclides (fission products and actinides), the total spontaneous fission rate, total photon emission rate, activity and decay heat at 789,406 different BU, CT, IE configurations for UO₂ fuel and at 531,991 different BU, CT, IPC configurations for MOX fuel. The fuel library is organized in a publicly available comma separated value file, thus its further analysis is possible and simple.

Data from calorimetric decay heat measurements of five used PWR 17x17 nuclear fuel assemblies

Raw data from calorimetric measurements of five nuclear fuel assemblies of the PWR 17 × 17 type are provided. Measurements of the temperature both inside a calorimeter, in which the fuel assembly was placed, as well as outside, were performed as a function of time while water circulating inside the calorimeter heats up from radiation emitted in the radioactive decay of material in the fuel assembly. The data contain also measurements of dose rate in the water outside the calorimeter. Data from 38 measurements using an electrically heated model of a fuel assembly are also provided to be used for, e.g., calibration.

The data can be used for validation of computer codes for modelling of nuclear systems, e.g. nuclear reactors, storage and transport of nuclear fuel or systems for geological disposal.

Optimization of database for identification need of unknown spent nuclear fuel samples

Recent years, the work of nuclear forensics has been greatly promoted in many aspects, and an important work is the establishment and application of nuclear forensic database. For now, this research is mainly based on uranium ore, because there is plenty of uranium ore information that is open and is easy to form database available for attribution. However, as the potential threat of spent nuclear fuel gets more and more attention, the number of researches on the identification of spent nuclear fuel via database is increasing. Since there is no public spent nuclear fuel database that is proper for attribution, such kind of work is mainly on methodology study. This paper focuses on the use of database for the identification of spent nuclear fuel. A database is first constructed with numerical simulation results and is used to identify samples both from simulation and experimental measurements to study the availability and applicability. Then samples from real database are used to optimize the database constructed with simulation result to better meet the need of real nuclear forensics scenarios.

Peroxide-treated metal-organic framework templated adsorbents for remediation of high level nuclear waste

Remediation of legacy nuclear waste is one of the greatest challenges faced by the US Department of Energy, with projected cleanup efforts requiring over five decades and hundreds of billions of dollars. New materials are necessary to accelerate waste processing, achieving time and financial savings. Herein we report a peroxide treatment to a Ti metal-organic framework (MOF) and related MOF-templated adsorbents. The resulting materials displayed exceptional affinity for Am(III), achieving distribution coefficients in excess of 10⁵ mL/g, and out-performing state-of-the-art benchmarks monosodium titanate (MST) and peroxo-treated modified MST (mMST) for removal of ⁸⁵Sr(II) and ^{239, 240}Pu(IV) from legacy nuclear waste simulant.

Determination of 107Pd in Pd purified by selective precipitation from spent nuclear fuel by laser ablation ICP-MS

Determination of radiopalladium ¹⁰⁷Pd is required to ensure radiation safety of the Pd extracted from spent nuclear fuel for recycling or disposal. We employed nanosecond laser ablation inductively coupled plasma quadrupole mass spectrometry (ns-LA-ICP-QMS) to simplify the analytical procedure of ¹⁰⁷Pd. Pd was separated through a selective Pd precipitation reaction induced by pulsed laser irradiation that reduces Pd(II) ions to metal Pd(0). Laser ablation facilitates direct measurement of the Pd precipitates, skipping the dissolution and dilution procedure with aqua regia and HCl, which causes serious corrosion damage to the introduction system of the ICP. In the present study, ¹⁰²Pd in natural Pd standard solution was used as an internal standard owing to its absence in spent nuclear fuel. Pd precipitates with diameters ranging from 0.2 to 0.5 μm, obtained by pulsed laser irradiation, were embedded uniformly on the surface of the centrifugal filter to form a microscopically thin and flat Pd surface. The resulting homogeneous Pd layer is suitable for obtaining a stable signal ratio of ¹⁰⁷Pd/¹⁰²Pd (< 4%, 2RSD). The mass bias-corrected ratio of ¹⁰⁷Pd/¹⁰²Pd and the amount of ¹⁰⁷Pd were 0.163 ± 0.004 and 17.8 ± 0.6 ng, respectively, which correspond to the values obtained by solution nebulization measurement after the dissolution of identical Pd precipitates. Graphical abstract ᅟ.

Consistent criticality and radiation studies of Swiss spent nuclear fuel: The CS2M approach

In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS₂M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO₂ assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources.

Dissolution of studtite [UO2(O2)(H2O)4] in various geochemical conditions

This study determined the dissolution rate of studtite, (UO₂)O₂(H₂O)₄, which can be formed by reaction between H₂O₂ and UO₂²⁺ that leaks from spent nuclear fuel (SNF) in deep geological repositories. The batch dissolution experiments were conducted using synthesized studtite under different solution conditions with varying pHs and concentrations of HCO₃^- and [H₂O₂] in synthetic groundwater. The experimental results suggested that carbonate ligand and H₂O₂ in groundwater accelerated the dissolution of studtite and uranium (U) release. Above 10^-5 M of H₂O₂ initial concentration, the released uranium concentration in solution decreased, possibly as a result of reprecipitation of studtite due to reaction between uranium and H₂O₂. The results will be useful to assess the comprehensive transport of uranium from both nuclear waste and SNF stored in deep geological repositories.

Retention of cesium and strontium by uranophane, Ca(UO2)2(SiO3OH)2·5H2O

This work determines the capacity of uranophane, one of the long-term uranyl secondary solid phases formed on the spent nuclear fuel (SNF), to retain radionuclides (cesium and strontium) released during the dissolution of the SNF. Sorption was fast in both cases, and uranophane had a high sorption capacity for both radionuclides (maximum sorption capacities of 1.53·10^-5 mol m^-2 for cesium and 3.45·10^-3 mol m^-2 for strontium). The high sorption capacity of uranophane highlights the importance of the formation of uranyl silicates as secondary phases during the SNF dissolution, especially in retaining the release of radionuclides not retarded by other mechanisms such as precipitation.

Isolation and characterization of culturable bacteria present in the spent nuclear fuel pool water

A spent nuclear fuel (SNF) pool is a key facility for safe management of nuclear waste, where spent nuclear fuel rods are stored in a water pool. The spent fuel rods carry a significant amount of radioactivity; they are either recycled or stored for further processing. Pool water acts as a heat sink as well as a shield against the radiation present in spent/burned fuel rods. The water used in these pools is filtered by an ultra-filtration process which makes certain the purity of water. As the life span of these pools is approximately 20 to 40 years, the maintenance of pure water is a big challenge. A number of researchers have shown the presence of bacterial communities in this ultrapure water. The bacterial types present in SNF pool water is of increasing interest for their potential bioremediation applications for radioactive waste. The present study showed the isolation of six bacterial species in the SNF pool water samples, which had significant radio-tolerance (D₁₀ value 248 Gy to 2 kGy) and also biofilm-forming capabilities. These strains were also investigated for their heavy metal removal capacity. Maximum biofilm-mediated heavy metal (Co and Ni) removal (up to 3.8 μg/mg of biomass) was observed by three isolates (FPB1, FPB4, and FPB6). The ability of these bacterial isolates to survive in radioactive environments can be of great interest for remediation of radioactive contaminants.

A quantitative model of water radiolysis and chemical production rates near radionuclide-containing solids

We present a mathematical model that quantifies the rate of water radiolysis near radionuclide-containing solids. Our model incorporates the radioactivity of the solid along with the energies and attenuation properties for alpha (α), beta (β), and gamma (γ) radiation to calculate volume normalized dose rate profiles. In the model, these dose rate profiles are then used to calculate radiolytic hydrogen (H2) and hydrogen peroxide (H2O2) production rates as a function of distance from the solid-water interface. It expands on previous water radiolysis models by incorporating planar or cylindrical solid-water interfaces and by explicitly including γ radiation in dose rate calculations. To illustrate our model's utility, we quantify radiolytic H2 and H2O2 production rates surrounding spent nuclear fuel under different conditions (at 20 years and 1000 years of storage, as well as before and after barrier failure). These examples demonstrate the extent to which α, β and γ radiation contributes to total absorbed dose rate and radiolytic production rates. The different cases also illustrate how H2 and H2O2 yields depend on initial composition, shielding and age of the solid. In this way, the examples demonstrate the importance of including all three types of radiation in a general model of total radiolytic production rates.