Modeled salt density for nuclear material estimation in the treatment of spent nuclear fuel

Spent metallic nuclear fuel is being treated in a pyrometallurgical process that includes electrorefining the uranium metal in molten eutectic LiCl-KCl as the supporting electrolyte. We report a model for determining the density of the molten salt. Material balances account for the net mass of salt and for the mass of actinides present. It was necessary to know the molten salt density, but difficult to measure. It was also decided to model the salt density for the initial treatment operations. The model assumes that volumes are additive for the ideal molten salt solution as a starting point; subsequently, a correction factor for the lanthanides and actinides was developed. After applying the correction factor, the percent difference between the net salt mass in the electrorefiner and the resulting modeled salt mass decreased from more than 4.0% to approximately 0.1%. As a result, there is no need to measure the salt density at 500 °C for inventory operations; the model for the salt density is found to be accurate.     

Thermal criteria for terminal storage of spent nuclear fuel assemblies in bedded salt

Design considerations, particularly thermal design considerations, for terminal storage of spent nuclear fuel in a bedded salt repository are discussed in a set of four papers which address: thermal criteria, uncoupling thermal problems, near-field temperatures, and optimization of mine arrangement. This paper outlines the thermal criteria used in the conceptual design of a bedded salt repository and discusses the historical rationale behind developments of each criterion.     

Recovery of minor actinides from spent molten salt waste and decontamination of molten salt waste

Recovery of minor actinides from spent molten salt is one of the important issues. Decontamination of spent molten salt waste is also the problem to be solved for establishment of pyrochemical reprocessing. The decontamination method of spent molten salt waste with recovery of minor actinides has been proposed. Our proposed process is based on the hydrometallurgical process. This process consists of the following processes. First, the spent molten salt waste is dissolved in aqueous solution. Next, the minor actinides are recovered by chromatographic techniques using the pyridine resin in the methanolic hydrochloric acid solution. In the last process, the spent molten salt waste is decontaminated by the cation-exchange chromatography. In the present paper, the adsorption behavior of minor actinides, rare earth elements, alkaline earth elements, and alkali metal elements on pyridine resin is reported. The demonstration experiment of the recovery of the minor actinides from simulant spent molten salt waste is also reported.     

Tokamak based hybrid systems for fuel production and recovery from spent nuclear fuel

The effect of the internal elements of a tokamak with different coolants on the transmutation rate of long-lived actinides is studied. The Monte Carlo method is used to calculate nuclear reactions in a homogeneous model of a blanket and in specific designs of a blanket. The neutron-physical calculations of a homogeneous model of a blanket showed that the absorption of neutrons by the central column of the tokamak and their moderation by the beryllium coating slow the transmutation rate to 35% of the initial value with the most efficient utilization of lead coolant. For water coolant, this effect is negligible. In a heterogeneous model of a blanket where water coolant is used, plutonium must be added to the actinides (50%/50%). The use of lead as the coolant will increase the transmutation rate of the actinides without using plutonium. In this case, it will be possible to reprocess spent nuclear fuel from more than 10 VVER-1000 in the JUST-T hybrid reactor.     

Decontamination of Zircaloy cladding hulls from spent nuclear fuel

The feasibility of decontaminating spent fuel cladding hulls using hydrofluoric acid (HF) was investigated as part of the Global Energy Nuclear Partnership (GNEP) Separations Campaign. The concentrations of the fission product and transuranic (TRU) isotopes in the decontaminated hulls were compared to the limits for determining the low level waste (LLW) classification in the United States (US). The ⁹⁰Sr and ¹³⁷Cs concentrations met the disposal criteria for a Class C LLW; although, in a number of experiments the criteria for disposal as a Class B LLW were met. The TRU concentration in the hulls generally exceeded the Class C LLW limit by at least an order of magnitude. The concentration decreased sharply as the initial 30-40 μm of the cladding hull surface were removed. At depths beyond this point, the TRU activity remained relatively constant, well above the Class C limit.     

Characterization of alloy particles extracted from spent nuclear fuel

We characterized, for the first time, submicro- and nanosized fission product-alloy particles that were extracted nondestructively from spent nuclear fuel, in terms of noble metal (Mo–Ru–Tc–Rh–Pd–Te) composition, atomic level homogeneity and lattice parameters. The evidences obtained in this work contribute to an improved understanding of the redox chemistry of radionuclides in nuclear waste repository environments and, in particular, of the catalytic properties of these unique metal alloy particles.     

Recycling of uranium from spent RBMK fuel in the nuclear fuel cycle

It is shown that there is promise in using the uranium product obtained by reprocessing spent nuclear fuel from RBMK reactors as a non-initial fuel source for thermal reactors. A technical path for spent nuclear fuel from RBMK reactors is proposed: radiochemical reprocessing and obtaining oxides of recycled uranium. Oxides of the category RBMK-poor are packed and then stored in a near-surface storage facility; oxides of the category RBMK-rich are fluoridated, and UF₆ is fed into separation production for additional enrichment to the required content of ²³⁵U. Additional advantages of recycled RBMK uranium as a source of non-initial ²³⁵U are the low content of ²³²U and the relatively low activity of spent fuel, which simplifies its reprocessing.     

Uncoupling thermal problems of terminal storage of spent nuclear fuel assemblies in bedded salt

Spent nuclear fuel assemblies stored in bedded salt can be modeled with a large array of dimensioned decay heat sources (spent fuel assemblies) in an extended thermal conducting media. Although a finite-difference or finite-element representation of the total storage facility could be established, regions of the repository should be analyzed separately since a model of the total repository would require formidable digital computing capacity. This paper explains the basis for thermally analyzing the total storage facility with separate models for the stored fuel assembly package and the salt medium. In addition, the effect of fuel assembly packaging on the maximum fuel temperature, the related problems of fuel handling prior to storage, and uncoupling of the effects of mine ventilation and conduction in the salt medium are discussed.     

Validation of analytical methods for nuclear spent fuel reprocessing

For more than 50 years, CETAMA, the Commission for establishment of analytical methods from the French Alternative Energies and Atomic Energy Commission, has provided Certified Reference Materials and Interlaboratory Comparisons for the development and validation of analytical methods in the nuclear field. In the future, the nuclear spent fuel reprocessing industry will require new standards and methods to comply with high content plutonium fuel and new extraction solvents. These standards and methods will have to be fully validated in order to ensure the quality of the analytical results obtained by the laboratories.In this context, a new ²⁴²Pu reference material, certified for its isotopic composition, has been recently produced. A novel statistical approach for data processing has been used and has led to a certified value of 0.985459 ± 0.000052 for the n(²⁴²Pu)/n(Pu) atomic ratio. In addition, an interlaboratory comparison has also been organized for the validation of a method for the analysis of DMDOHEMA, and its degradation products. This compound is considered as a new extractant candidate in the frame of separation processes for transmutation of long-lived radionuclides. The methodology and results obtained in both cases are presented.     

Uranium measurements using laser-induced breakdown spectroscopy in lithium chloride-potassium chloride salt of pyroprocessing

In pyroprocessing,uranium(U) is recovered from molten LiCl-KCl salt,and,for safeguard purposes,it is important to analyze the U and Plutonium(Pu) concentrations in a timely manner.In the present work,salt samples containing U were fabricated.The laser used in the present work was an Nd:YAG laser with a wavelength of 532 nm,a laser energy on the sample of11.5 mJ,and a pulse repetition rate of 10 Hz.The plasma emission light was measured with an Echelle spectrometer.A total of 100 points on the sample surface were measured as the laser incident position was changed.The U and potassium(K) peaks in the spectrum were identified.Univariate and multivariate analyzes were conducted to determine the accuracy and limit of detection(LOD) of the laser-induced breakdown spectroscopy.     

10-MWt固态钍基熔盐堆乏燃料贮存系统临界安全影响分析

10-MWt固态钍基熔盐堆(Thorium-based Molten Salt Reactor-Solid Fuel,TMSR-SF)使用TRISO(Tri-structural isotropic)颗粒燃料元件,并采用熔融氟盐作为一回路冷却剂,附着在燃料元件上的熔盐有可能影响系统反应性。因此,需要分析在燃料元件的贮存过程中熔盐附着燃料元件对贮存临界安全的影响。使用SCALE6.1的TRITON(Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)模块对TMSR-SF堆芯建模并进行燃耗计算,使用MCNP对乏燃料贮存系统进行临界计算。分别考虑熔盐浸渗球形燃料元件和熔盐包覆在球形燃料元件表面两种典型情况下,熔盐附着对贮存系统反应性的影响。针对乏燃料贮存系统,以浸渗最大量,即熔盐体积是石墨体积的13.9%为前提,临界计算结果表明,熔盐浸渗入石墨基体贮存系统的反应性比熔盐包覆在球形燃料元件表面的贮存系统的反应性要大5%;与没有熔盐附着的情况相比,有熔盐附着的情况下贮存系统反应性要大15%。对乏燃料贮存系统的临界安全分析可知,两种典型的熔盐附着模型对贮存系统的反应性存在一定的影响,但无论是熔盐浸渗还是包覆,贮存系统仍处于次临界,意味着贮存系统在正常工况下是安全的。