Fuel Conditioning Facility Research Articles

A literature review of pyroprocessing safeguards

Pyroprocessing is a promising technology for reprocessing used nuclear fuel (UNF) from light water reactors (LWR) and sodium fast reactors (SFR). With the advancement of fast reactors with projects from companies such as TerraPower, the prospects of pyroprocessing are more promising than ever before, as fast reactors can consume all actinides, not just uranium and plutonium. Before pyroprocessing can be implemented commercially, pre-existing safeguards used for aqueous reprocessing must be adapted for pyroprocessing. Safeguards, as determined by both the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), are discussed in this review as guidelines for approaches used in both domestic and international reprocessing plants. The implementation of safeguards in aqueous reprocessing was then reviewed based on the experience from existing aqueous facilities. The experimental pyroprocessing facilities were identified as pyroprocessing plants in the design stage with one operating exception of the Fuel Conditioning Facility (FCF). The safeguard methods implemented or designed for each and the accompanying challenges of utilizing existing safeguards in pyroprocessing are considered and a summary of applicable approaches is included.

Options Study for the Neutralization of Elemental Sodium During the Pyroprocessing of Used Nuclear Fuel

An options study was performed for the treatment of residual elemental sodium in driver plenums following the chopping operation during the pyroprocessing of used nuclear fuel. Given the pending availability of a multi-function furnace for distillation and consolidation operations in the Fuel Conditioning Facility, the furnace was considered for the processing of driver plenums. Although two options (oxidation and distillation) could be performed in the multi-function furnace, neither option has been developed sufficiently to date to warrant the use of the furnace for treatment operations. Thus, it was decided to defer the treatment of elemental sodium from driver plenums in the multi-function furnace until more developed technologies and/or furnaces become available. In the interim, storage of the plenums and characterization efforts are recommended.

Fuel conditioning facility inert gas filled reprocessing hot cell leak rate measurement

The Idaho National Laboratory Fuel Conditioning Facility conducts treatment of U.S. Department of Energy owned metallic fast reactor spent fuel using pyro-processing. Certain steps in pyro-processing generate pyrophoric metal laden with radioactive material. Therefore, the process is conducted within an inert atmosphere to prevent ignition. The Fuel Conditioning Facility uses a 60,000ft3 (1700m3) shielded hot cell environment with a purified argon gas atmosphere. Significant leakage of air into the hot cell can raise the oxygen concentration to the point where exposed pyrophoric metal will burn, which can subsequently result in release of radioactive material through the air in-leakage point. Additionally, limited air in-leakage can cause exposed metal to oxidize, rendering it unfit for pyro-processing. To ensure hot cell confinement boundary integrity, the Fuel Conditioning Facility executes periodic surveillance of the confinement boundary leak rate. This paper describes the leak rate surveillance techniques associated with this large inert-gas-filled hot cell.Two methods have been developed to determine the cell leak rate. One method uses the cell atmosphere oxygen concentration change over time. The other method uses the cell pressure and temperature change over time. The oxygen rate increase method is a very accurate method of determining the air leak rate during many types of operating conditions. However, the data observation period must be sufficiently long to obtain an accurate value. The pressure-temperature method is not as time consuming but is significantly less accurate than the oxygen concentration method.The cell leak rate is extremely small, less than 0.01cfm, given the very large cell size. The oxygen concentration method is accurate to approximately ±0.002cfm. The pressure-temperature method is accurate to about ±1cfm. The pressure-temperature leak rate method can be used as a confirmation of the oxygen method results or as a backup if the oxygen monitors are not working.

Development and Optimization of Voltammetry for Real Time Analysis of Multi-Component Electrorefiner Salt

There is interest by several countries in pursuing pyroprocessing of spent nuclear fuel –including both nuclear weapons states and non-nuclear weapons states. This technology features electrorefiners that can be operated to recover fissile actinides from spent nuclear fuel. This separation capability can be viewed as a proliferation risk factor, which necessitates the implementation of nuclear safeguards. For non-nuclear weapons states, safeguarding would be the responsibility of the International Atomic Energy Agency. Unlike nuclear facilities in which items can be discretely tracked, inspecting and safeguarding pyroprocessing facilities requires complicated monitoring systems and material control and accountability approaches to support the safeguards system. Therefore, there needs to be developed high precision monitoring technology for electrorefining and other key unit operations in pyroprocessing that can be implemented in real or near real time. Currently in research and development facilities such as the Fuel Conditioning Facility at Idaho National Laboratory, they have been employing mass tracking simulation supplemented with destructive analysis for the last two decades. This approach meets neither the need for timeliness or accuracy needed to detect significant quantities of material diversion in commercial facilities. Since much of the inventory of actinides in electrorefiners are held up in the molten salt, electrochemical sensors would be an ideal solution to accurately monitor actinide concentrations in real time. Electrochemical analysis methods such as cyclic voltammetry (CV), chronoamperometry (CA), and chronopotentiometry (CP), and normal pulse voltammetry (NPV) can be used to develop correlations between electric signals measured and concentrations of various ions in the molten salt. In salts with multiple components and high overall concentration of active solute, one major difficulty in applying these electroanalytical methods is to prevent area growth of the working electrode (WE)—which is largely unpredictable and affects the observable current density. Practically, all of the aforementioned voltammetry methods can encounter this problem. However, during an NPV scan, the metal deposited onto the WE is periodically oxidized back into the salt, which should effectively minimize working electrode area growth. NPV was, thus, the main focus of our study. Pulse time, relaxation time, and relaxation potential were all varied in an attempt to isolate the diffusion-limited partial current values for the different active ions in the salt. CV and CA were also applied to supplement and validate the NPV results. CV was employed to determine the potentials at which different ions reduce as well as the relaxation potential used to clean off deposited metals from the WE. CA was utilized to validate if the pulse time selected was still within the time period that the change in current was described by the Cottrell equation. Therefore, a method that was a combination of NPV, CV, and CA was applied to develop and optimize real time analysis of electrorefiner salt. Two salt mixtures were investigated in this study, LiCl-KCl-UCl3-MgCl2 and LiCl-KCl-UCl3-GdCl3, where Mg was the surrogate for Pu and Gd represented the rare earth elements in the electrorefiner salt based on the similarity in standard reduction potentials. While previously, correlations between limiting current and concentration in molten salt had been reported only up to about 1.7 wt% UCl3, this study included limiting current and concentration correlations for binary systems with up to 9 wt% UCl3. The highest MgCl2 concentration tested was 1.5 wt%, which represented 5.5% PuCl3 because the limiting current is proportional to molar concentration rather than weight percentage. When comparing limiting current density with MgCl2 concentration at fixed 1 wt% UCl3, the results did not show a good linear fit, as the slope increased with each concentration increment. However, R2 of 0.9996 was obtained when the data was fitted with a second order polynomial. The nonlinear results show that there was possibly a slight WE surface area increase. Also it is possible that the diffusion coefficient increased significantly with high concentration, resulting in higher current density. Further investigation with shorter pulse time is required to validate these hypothesizes. Nevertheless, the results show that electrochemical sensors would be an ideal solution to accurately monitor electorefiner salt concentrations in real time.

Transient response and radiation dose estimates for breaches to a spent fuel processing facility

This paper describes the analysis of the design basis accident for Idaho National Laboratory Fuel Conditioning Facility (FCF). The facility is used to process spent metallic nuclear fuel. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

ELECTROCHEMICAL PROCESSING OF USED NUCLEAR FUEL

As part of the Department of Energy’s Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

A Computational Model of the Mark-IV Electrorefiner: Phase I—Fuel Basket/Salt Interface

Spent driver fuel from the Experimental Breeder Reactor-II is currently being treated in the Mark-IV electrorefiner in the Fuel Conditioning Facility at Idaho National Laboratory. The modeling approach to be presented here has been developed to help understand the effect of different parameters on the dynamics of this system. The first phase of this new modeling approach focuses on the fuel basket/salt interface involving the transport of various species found in the driver fuels (e.g., uranium and zirconium). This approach minimizes the guessed parameters to only one, the exchange current density (i0). U3+ and Zr4+ were the only species used for the current study. The result reveals that most of the total cell current is used for the oxidation of uranium, with little being used by zirconium. The dimensionless approach shows that the total potential is a strong function of i0 and a weak function of wt % of uranium in the salt system for initiation processes.

Fuel Conditioning Facility Electrorefiner Model Predictions versus Measurements

Electrometallurgical treatment of spent nuclear fuel is performed in the Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) by electrochemically separating uranium from the fission products and structural materials in a vessel called an electrorefiner (ER). To continue processing without waiting for sample analyses to assess process conditions, an ER process model predicts the composition of the ER inventory and effluent streams via multi-component, multi-phase chemical equilibrium for chemical reactions and a numerical solution to differential equations for electro-chemical transport. The results of the process model were compared to the electrorefiner measured data.

Engineering-Scale Liquid Cadmium Cathode Experiments

Recovery of transuranic actinides (TRU) using electrorefining is a process being investigated as part of the Department of Energy (DOE) Advanced Fuel Cycle Initiative (AFCI). TRU recovery via electrorefining onto a solid cathode is very difficult as the thermodynamic properties of transuranics are not favourable for them to remain in the metal phase while significant quantities of uranium trichloride exist in the electrolyte. Theoretically, the concentration of transuranics in the electrolyte must be approximately 106 greater than the uranium concentration in the electrolyte to produce a transuranic deposit on a solid cathode. Using liquid cadmium as a cathode contained within a LiCl-KCl eutectic salt, the co-deposition of uranium and transuranics is feasible because the activity of the transuranics in liquid cadmium is very small. Depositing transuranics and uranium in a liquid cadmium cathode (LCC) theoretically requires the concentration of transuranics to be two to three times the uranium concentration in the electrolyte. Three LCC experiments were performed in an Engineering scale elecdtrorefiner, which is located in the argon hot cell of the Fuel Conditioning Facility at the Materials and Fuels Complex on the Idaho National Laboratory. Figure 1 contains photographs of the LCC assembly in the hot cell prior to the experiment and a cadmium ingot produced after the first LCC test. Figure 1. Liquid Cadmium Cathode (left) and Cadmium Ingot (right) The primary goal of the engineering-scale liquid cadmium cathode experiments was to electrochemically collect kilogram quantities of uranium and plutonium via a LCC. The secondary goal was to examine fission product contaminations in the materials collected by the LCC. Each LCC experiment used chopped spent nuclear fuel from the blanket region of the Experimental Breeder Reactor II loaded into steel baskets as the anode with the LCC containing 26 kg of cadmium metal. In each experiment, between one and two kilograms of heavy metal was collected in the LCC after passing an integrated current over 500 amp hours. Analysis of samples from the liquid cadmium cathode ingots showed detectable amounts of transuranics and rare-earth elements. Acknowledgements K. B. Davies and D. M. Pace for the mechanical and electrical engineering needed to prepare the equipment for the engineering-scale liquid cadmium cathode experiments.

Fuel Conditioning Facility Electrorefiner Process Model

The Fuel Conditioning Facility at the Idaho National Laboratory processes spent nuclear fuel from the Experimental Breeder Reactor II using electro‐metallurgical treatment. To process fuel without waiting for periodic sample analyses to assess process conditions, an electrorefiner process model predicts the composition of the electrorefiner inventory and effluent streams. For the chemical equilibrium portion of the model, the two common methods for solving chemical equilibrium problems, stoichiometric and non‐stoichiometric, were investigated. In conclusion, the stoichiometric method produced equilibrium compositions close to the measured results whereas the non‐stoichiometric method did not.

Progress in Electrometallurgical Treatment of Spent Nuclear Fuel

Argonne National Laboratory (ANL) has been developing electrometallurgical treatment processes since the late 1980s. These non-aqueous processes separate uranium from fission products via a series of high temperature steps: electrorefining, cathode processing, and metal casting. During the past five years, more than 1.8 tonnes of spent nuclear fuel have been processed in the Fuel Conditioning Facility at ANL-West. This operating experience has provided the basis for a more detailed evaluation of these processes. High-level waste forms have been produced at a demonstration scale with the materials from the spent fuel. Although processing is currently limited to 600 kilograms per year by funding constraints, plans and technical advances are being implemented to increase processing to 5 tonnes per year. These operations include producing high-level waste for final geologic disposal.

Operational analysis and improvement of a spent nuclear fuel handling and treatment facility using discrete event simulation

Spent nuclear fuel handling and treatment often require facilities with a high level of operational complexity. Simulation models can reveal undesirable characteristics and production problems before they become readily apparent during system operations. The value of this approach is illustrated here through an operational study, using discrete event modeling techniques, to analyze the Fuel Conditioning Facility at Argonne National Laboratory and to identify enhanced nuclear waste treatment configurations. The modeling approach and results of what-if studies are discussed. An example on how to improve productivity is presented.

Testing hot cell shielding in the fuel conditioning facility.

A comprehensive shield test program for a hot cell complex, the Fuel Conditioning Facility at Argonne National Laboratory, has been completed with minimum radiation exposure to participants. The recently modified shielding design for two hot cells and their associated transfer paths for irradiated materials was analyzed and tested for attenuating gamma rays from mixed fission product sources. Testing was accomplished using 0.37 TBq (10 Ci) and 518 TBq (14,000 Ci) 60Co sources. Of specific concern were radiation levels around wall penetrations and the interface between transport casks and the cell floor. Detailed measurements were made for surfaces that bound the hot cells, a transfer tunnel between the two cells, and storage pits that extend below the floors of both cells. In addition to surface measurements, dose equivalent rates in adjacent corridors were determined when the larger source was exposed. Results indicate that with some administrative controls, the facility shields are adequate to meet the design criterion that limits annual dose to less than 10 mSv (1 rem) for facility workers.

Testing shields in the Argonne National Laboratory fuel conditioning facility support areas.

Testing has been completed for two lightly shielded areas that support operations in the Fuel Conditioning Facility at the Argonne National Laboratory site at the Idaho National Engineering Laboratory. Operational requirements dictated the use of a radiography source containing 0.44 TBq (12 Ci) of 192Ir to challenge reinforced concrete and steel shields that surround a decontamination, maintenance, and repair area for contaminated equipment used in hot cell operations. A more intense source containing 0.89 TBq (24 Ci) of 192Ir was used to test lead shot and steel shields around tanks in a radioactive liquid waste system and the boundaries of the room that contained it. Measurement procedures were developed to find design flaws and construction deficiencies while minimizing radiation exposure to test participants. While the shields are adequate to limit gamma ray deep dose equivalents to 10 mSv y(-1) (1 rem y(-1)) or less to facility personnel, several modifications were necessary to assure that the attenuation is adequate to keep dose rates less than 5 microSv h(-1) (0.5 mrem h(-1)) in normally occupied areas.