Molten salt synthesis of Ce doped zirconolite for the immobilisation of pyroprocessing wastes and separated plutonium
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99 GENERAL AND MISCELLANEOUS↗
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Immobilization of LiCl-Li2O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods
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Riley, Brian J. (ORCID:0000000277456730)↗
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.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
An approach to developing an integrated pyroprocessing simulator
Pyroprocessing has been studied for a decade as one of the promising fuel recycling options in Korea. We have built a pyroprocessing integrated inactive demonstration facility (PRIDE) to assess the feasibility of integrated pyroprocessing technology and scale-up issues of the processing equipment. Even though such facility cannot be replaced with a real integrated facility using spent nuclear fuel (SF), many insights can be obtained in terms of the world's largest integrated pyroprocessing operation. In order to complement or overcome such limited test-based research, a pyroprocessing Modelling and simulation study began in 2011. The Korea Atomic Energy Research Institute (KAERI) suggested a Modelling architecture for the development of a multi-purpose pyroprocessing simulator consisting of three-tiered models: unit process, operation, and plant-level-model. The unit process model can be addressed using governing equations or empirical equations as a continuous system (CS). In contrast, the operation model describes the operational behaviors as a discrete event system (DES). The plant-level model is an integrated model of the unit process and an operation model with various analysis modules. An interface with different systems, the incorporation of different codes, a process-centered database design, and a dynamic material flow are discussed as necessary components for building a framework of the plant-level model. As a sample model that contains methods decoding the above engineering issues was thoroughly reviewed, the architecture for building the plant-level-model was verified. By analyzing a process and operation-combined model, we showed that the suggested approach is effective for comprehensively understanding an integrated dynamic material flow. This paper addressed the current status of the pyroprocessing Modelling and simulation activity at KAERI, and also predicted its path forward.
42 ENGINEERING↗
Seismoacoustic Monitoring of Pyroprocessing Equipment
With the projected increase in world nuclear capacity comes the hurdle of spent nuclear fuel. Pyroprocessing is one method to process irradiated fuel by using high temperatures and electrochemical steps to separate radioactive components. High temperatures, high radiation levels, and equipment confined to a heavily shielded hot cell are a few of the challenges introduced regarding safeguards for pyroprocessing. The first stage of the process utilizes an element chopper to cut irradiated fuel into smaller pieces, offering the potential to identify equipment operation through seismoacoustic monitoring. Eleven seismic and infrasound sensors were deployed at various locations and distances near a pyroprocessing hot cell to evaluate signals emitted from the chopper. Signals were processed using short-time Fourier transforms and manually scanned to locate chopping events. Operator logs were then collected for a ground truth comparison. Results indicate that both infrasound and seismic signals offer the ability to accurately measure this stage of pyroprocessing, though seismic signals are more prominent. Sensor locations are also evaluated to determine where the signal is detectable. This type of monitoring offers a method to aid in safeguards and proliferation detection by validating a pyroprocessing facility’s schedule and identifying activity that does not align with records.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Safeguards in Pyroprocessing: an Integrated Model Development and Measurement Data Analysis
Pyroprocessing is an electrochemical method based on the molten salt electrolyte, mainly the LiCl-KCl eutectic molten salt, to recycle the used nuclear fuel. For a conceptual design of commercial pyroprocessing facility, tons of special nuclear materials, namely U and Pu, may be involved, which could be used for non-peaceful purposes if they are diverted. Effective safeguards approaches have to be developed prior to the development and construction of a pyroprocessing facility. Present research focused on two main objectives, namely calculating the properties of nuclear species in LiCl-KCl molten salt and developing integrated model to safeguard a pyroprocessing facility. Understanding the characteristics of special nuclear materials in LiCl-KCl eutectic salt is extremely important to understand their behaviors in an electrorefiner. The model development for the separation processes in the pyroprocessing, including electrorefining, actinide drawdown, and rare earth drawdown benefits the understanding of material transport and separation performance of these processes under various conditions. The output signals, such as potential, current, and species concentration contribute to the material balance closure and provide safeguards signatures to detect the scenarios of diversion. U and Pu are the two main elements concerned in this study due to our interest in safeguards.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Modeling of Pyroprocessing Hot Cell for Process Monitoring Evaluation
In this report, we examine studies undertaken to determine the applicability of applying measurements with a neutron counter within a pyroprocessing hot cell. The studies focus on the modeling of the hot cell and neutron counter in Monte Carlo N- Particle Code (MCNP) to determine expected radiation effects and applicability of neutron counters. The neutron counter of focus was the High Dose Neutron Detector (HDND) developed at Los Alamos National Laboratory (LANL). We studied the HDND to determine what would be the detector response under normal background conditions with operating process equipment as well as the background’s potential effect on counting applications involving a U/TRU ingot produced by the pyroprocessing electrorefiner. To perform this task, we were required to develop models of the pyroprocessing process cell and unit operations and generate source terms for various different times in the process. The data for the pyroprocessing materials came from SNL’s Safeguards and Security Process Model (SSPM) as part of MPACT’s advanced integration mission. This report will discuss the development of the pyroprocessing process cell and unit operations models in MCNP as well as various simulations to determine the applicability of the HDND for process monitoring efforts in the hot cell.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Cesium Removal from Surrogate Pyroprocessing Salt by Electrodeposition
Active metals in used nuclear fuel dissolve into the salt during pyroprocessing and are not removed by electrorefining or drawdown operations. The buildup of 137 Cs over time increases the heat load and ionizing radiation level of the salt such that it must be replaced frequently, resulting in a significant amount of salt waste. An effective means of managing cesium in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. A previous report summarized issues that must be addressed when developing a removal strategy and assessed the suitability of existing methods and remaining technological gaps to their application (Rose and Thomas 2023). Cesium is extremely stable in molten salt as a chloride—even more stable than the LiCl-KCl eutectic base salt used for pyroprocessing fuel—which makes removing cesium a challenge. However, sufficiently strong atomic interactions occur between active metal species and liquid metals that make the electrodeposition of active metal fission products into liquid metal electrodes energetically favorable. The feasibility of recovering cesium from LiCl-KCl pyroprocessing salt through electrodeposition into liquid metals is eing assessed by identifying potentially effective liquid metals and performing tests to determine the effectiveness of electrodepositing cesium from a LiCl/KCl salt into these liquid metals. Previous studies investigating the electrodeposition of Sr 2+ , and Ba 2+ into zinc, cadmium, bismuth, lead, tin and antimony have shown that alkali and alkaline earth metals can be electrodeposited at liquid metal cathodes (Kim et al., 2018). The removal of Ba 2+ and Sr 2+ was measured to be more efficient than the removal of monovalent cations due to the greater thermochemical driving force for alloying those elements with the liquid metal (Jang et al. 2022). Because the equilibrium potentials are dependent on the interactions of the active metal in the liquid metal, it is likely that the other alkali metals Li + , and K + , will deposit from LiCl/KCl salt with the Cs + . Therefore, application of this method to recover active metals from pyroprocessing salt will benefit from the use of a liquid metal and set of operating conditions that sufficiently increase the reduction potential of cesium to remove cesium from the waste salt with an acceptable amount of co-deposited lithium and potassium.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
3S (Safeguards, Security, Safety) based pyroprocessing facility safety evaluation plan
The big advantage of pyroprocessing for the management of spent fuels against the conventional reprocessing technologies lies in its proliferation resistance since the pure plutonium cannot be separated from the spent fuel. The extracted materials can be directly used as metal fuel in a fast reactor, and pyroprocessing reduces drastically the volume and heat load of the spent fuel. KAERI has implemented the SBD (Safeguards-By-Design) concept in nuclear fuel cycle facilities. The goal of SBD is to integrate international safeguards into the entire facility design process since the very beginning of the design phase. This paper presents a safety evaluation plan using a conceptual design of a reference pyroprocessing facility, in which 3S (Safeguards, Security, Safety)-By-Design (3SBD) concept is integrated from early conceptual design phase. The purpose of this paper is to establish an advanced pyroprocessing hot cell facility design concept based on 3SBD for the successful realization of pyroprocessing technology with enhanced safety and proliferation resistance.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Acoustic Monitoring of Pyroprocessing Equipment
This paper provides an introduction to using acoustic monitoring to advance detection techniques for pyroprocessing in support of nuclear safeguards and non-proliferation. The usage of free air acoustic monitoring has been previously demonstrated at Idaho National Laboratory (INL) facilities such as the Advanced Test Reactor and the National Security Test Range. However, the proposed work revolves around a new deployment environment, the Fuel Conditioning Facility, that brings forward several questions regarding the performance of the technology in non-free air media. The confinement of the pyroprocessing equipment to a heavily shielded hot cell, the atmosphere of the hot cell containing argon gas, and the radiation dose inside the hot cell are all new environments for acoustic monitoring. To our knowledge, acoustic measurements have not been completed in such an environment before. This offers a new opportunity to study not only the acoustic signatures of the equipment inside of the hot cell, but also the propagation of the signals through the hot cell and at distances away from their origination. The objective of this paper is to explain the planned instruments to monitor the Fuel Conditioning Facility in order to evaluate acoustic signals emitted from equipment during various stages of operation. Identifying these signals can potentially enable the identification of specific pieces of equipment used in pyroprocessing and produce information of their operational status. If successful, this type of monitoring could offer a new method to aid in safeguards and proliferation detection of pyroprocessing activities.
98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
Uncertainty improvement of 22 Na based radioactive tracer dilution for determining total mass of pyroprocessing molten salt systems by 154 Eu removal
To determine the total salt mass of the molten salt systems for pyroprocessing spent nuclear fuels, a 22 Na based radioactive tracer dilution was studied in Idaho National Laboratory in recent years. This 22 Na based RTD technique was deemed feasible, but due to the gamma energy peak of 22 Na coinciding with one of the energy peaks of 154 Eu radioisotope in the molten salt, the uncertainty of the 22 Na radioactivity in the 22 Na-spiked salt samples was quite high. To improve the uncertainty of the 22 Na based RTD technique, we proposed to chemically remove the 154 Eu of the salt samples by DGA resin for gamma spectroscopy. The effectiveness of removing 154 Eu on uncertainty improvement was evaluated. Furthermore, it was found that (1) the 154 Eu fission product effect on the uncertainty and detection limit can be effectively eliminated by chemically removing the 154 Eu during the salt sample preparation and (2) the uncertainty of 22 Na radioactivity in the salt samples for electrorefining was significantly improved from 13% to 2%, showing the potential of practical engineering application of 22 Na based RTD as a safeguards technique for molten salt systems for pyroprocessing spent nuclear fuels.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Conceptual Design of a Pilot-Scale Pyroprocessing Facility
Argonne National Laboratory and Merrick & Company introduced the conceptual design of a pilot-scale (100 T/year) pyroprocessing facility for the treatment of used fuel generated by commercial light water reactors and subsequent treatment of waste streams generated during the process. The primary purpose of this study was to perform sufficient engineering for the pilot facility conceptual design so that credible capital and operating cost estimates could be developed. Initial safety, safeguards, and security assessments were also completed to provide a detailed evaluation in these areas that can significantly affect both capital and operating costs. Electrorefining-based pyroprocessing resulted in a compact hot-cell facility with few process equipment systems. The process equipment and support systems were estimated to cost $93 million and the facility $305 million for a project total cost of $398 million. The annual operating cost was estimated at $53 million/year Scaling up to a commercial-scale (400 T/year) was also assessed and the capital cost was estimated at $911 million with annual operating cost of $90 million/year.
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Projected Salt Waste Production from a Commercial Pyroprocessing Facility
Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100 MT/year of fuel is processed. Salt waste estimates range from 8 to 20 MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing.
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Characterization of Delayed-Particle Emission Signatures for Pyroprocessing, Part 1: Fuel Assembly
An initial study was conducted with nondestructive assay analysis using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron and delayed- gamma emission signature monitors to characterize burnup and isotopic composition for pyroprocessing. The study consists of five parts. In Part 1, a generic Westinghouse fuel assembly model was modified to create models with 3% and 5% 235 U enrichments, 20, 30, 40, and 50 GWd/MTU burnups, and 3-, 5-, 10-, 20-, and 30-year cooling times. MCNP6 was used to create detail-rich delayed-particle source terms for fuel assembly. Delayed-gamma activity ratios were then developed for conventional 134 Cs/ 137 Cs, 134 Cs/ 154 Eu, and 154 Eu/ 137 Cs markers as a function of burnup and of actinide mass for elemental uranium, neptunium, plutonium, americium, and curium. In addition, data mining using photon tally data was employed to discover potential new delayed-gamma activity ratios. The data mining effort was streamlined by the creation and use of several utility codes to extract and process data. The study confirms behavior for burnup dependence of 134 Cs/ 137 Cs, 134 Cs/ 154 Eu, and 154 Eu/ 137 Cs markers. A process for the discovery of potential new delayed-gamma markers has been developed, and results are presented. The actinide and nonactinide inventories for the fuel assembly models are used as source materials for the electrorefining models. Parts 2–5 contain analysis of a pyrochemical separations model that includes an MCNP6 mockup of the INL Mark IV electrorefiner (see companion documents). Part 2 focuses on delayed-particle emissions from the electrorefiner with loaded fuel dissolution baskets before electrorefining. Part 3 investigates delayed-particle emissions from the electrorefiner following electrorefining. Part 4 presents delayed-particle emissions from other pyroprocessing units that contain material developed by the electrorefining process in Part 3. Part 5 parallels Part 4, except for a 10 wt% plutonium reduction.
98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
Pyroprocessing of Fast Flux Test Facility Nuclear Fuel
Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Proliferation Detection via Acoustic Monitoring of Pyroprocessing Equipment and Related Systems
In support of nuclear safeguards and nonproliferation, this effort is leveraging acoustic and seismic sensors to advance detection of pyroprocessing activities. Pyroprocessing, or electrochemical processing, is a method of separating irradiated nuclear fuel into its actinide and fission product components, enabling the reuse of fuel materials and reducing the radiotoxicity of the remaining waste. Verification and accountability in the process poses several challenges because of the high radiation environment and the presence of material holdup through the facility. Irregular or unaccounted-for equipment operation can signify diversion or misuse of critical materials. Acoustic and seismic sensors offer the capability to identify periods of equipment operation and help verify adherence to accountancy records.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Investigations for the Recycle of Pyroprocessed Uranium
Given the renewed interest in uranium from the pyroprocessing of used nuclear fuel in a molten salt system, the two biggest hurdles for marketing the uranium are radiation levels and transuranic content. A radiation level as low as possible is desired so that handling operations can be performed directly with the uranium. The transuranic content of the uranium will affect the subsequent waste streams generated and, thus also should be minimized. Although the pyroprocessing technology was originally developed without regard to radiation and transuranic levels, adaptations to the process have been considered to market the uranium. Process conditions have been varied during the distillation and casting cycles of the process with increasing temperature showing the largest effect on the reduction of radiation levels. Transuranic levels have been reduced significantly by incorporating a pre-step in the salt distillation operation to remove a majority of the salt prior to distillation.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Gamma-ray spectra analyses of molten salts in spent nuclear fuels pyroprocessing facilities for mass measurement
To evaluate radioactive tracer dilution and gamma spectroscopy as a safeguards technique for monitoring the mass of liquid salt in molten salt systems, the gamma spectroscopy data of salt samples from the pyroprocessing facilities in Idaho National Laboratory were acquired and analyzed. The primary focus is the uncertainty analysis of the gamma radioactivity of 154 Eu isotope in the molten salts. The facilities include an electrorefiner for pyroprocessing spent oxide fuels (SOF-ER) and Mark-IV ER for processing the metallic fuels from Experimental Breeder Reactor-II (EBR-II). The relative gamma radioactivity uncertainty was consistently at 3% for 137 Cs for both Mark-IV ER and SOF-ER salts. However, for 154 Eu, it was 8% for SOF-ER salt and 7% for Mark-IV ER salt. Here, the main reason for the higher uncertainty for 154 Eu is believed to be the lower counting statistics in gamma spectroscopy. To improve the gamma radioactivity uncertainty of isotope 154 Eu, a longer data acquisition time (12 h instead of 4 h) and a higher count rate gamma detector for gamma spectroscopy were tested and it was found that the uncertainty for 154 Eu was significantly improved from 7% or 8–3%.
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