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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Sister Rod Destructive Examinations (FY22) Appendix F2: Evaluation of Fuel Rod Fatigue During Spent Fuel Transportation

This report documents work performed under the Spent Fuel and Waste Disposition’s Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 2 Milestone M2SF-23OR010201024, “FY22 Report on ORNL Sibling Rod Testing Results,” within work package SF-23OR01020102 and is an update to the work reported in M2SF-22OR010201047, M2SF-21OR010201032, M2SF-19ORO010201026, and M2SF- 19OR010201028. As a part of DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rods’ condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This appendix documents an evaluation of the fatigue data to enhance the technical basis for post-storage transportation, handling, and subsequent disposition and to identify future testing needs for Phase 2 of the project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Corrosion-Resistant Coatings on Spent Nuclear Fuel Canisters to Mitigate and Repair Potential Stress Corrosion Cracking (FY23 Status)

This report summarizes the activities performed by Sandia National Laboratories in FY23 to identify and test coating materials for the prevention, mitigation, and/or repair of potential chloride-induced stress corrosion cracking in spent nuclear fuel dry storage canisters. This work continues efforts by Sandia National Laboratories that are summarized in previous reports from FY20 through FY22 on the same topic. In FY23, Sandia National Laboratories, in collaboration with five industry partners through a memorandum of understanding, evaluated the physical, mechanical, and corrosion-resistance properties of eight different coating systems. The evaluation included thermal and radiation environments relevant to various time periods of storage for spent nuclear fuel canisters. The coating systems include polymeric (polyetherketoneketone, modified polyimide/polyurea, modified phenolic resin, epoxy), organic/inorganic ceramic hybrids (silane-based polyurethane hybrid and a quasi-ceramic sol-gel polyurethane hybrid), and coatings utilizing a Zn-rich primer applied to stainless steel coupons. The results and implications of these tests are summarized in this report. These analyses will be used to identify the most effective coatings for potential use on spent nuclear fuel dry storage canisters and to identify specific needs for further optimization of coating technologies for application on spent nuclear fuel canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

SCALE 6.3 Validation: Spent Nuclear Fuel

This report is the fifth volume in a series documenting the validation of SCALE 6.3 with ENDF/B-VII.1 libraries for nuclear criticality safety, reactor physics, radiation shielding, and spent nuclear fuel applications. This fifth volume, which focuses on validating SCALE capabilities that impact spent nuclear fuel applications, provides an update of the similar validation reported for SCALE 6.2.4. The experimental data used as basis for validation herein consist of measurement data for nuclide inventories and decay heat, including the following: 1. radiochemical assay (RCA) measurements of nuclides important to burnup credit, decay heat, and radiation shielding in 205 light-water reactor (LWR) spent nuclear fuel samples that cover burnups ranging up to 80 GWd/MTU and initial fuel enrichments up to 4.9% 235U; 2. full-assembly decay heat measurements for 236 LWR assemblies with initial fuel enrichments ranging up to 4% 235 U, assembly burnups of 5–51 GWd/MTU, and decay times after fuel discharge in the 2- to 27-year range (of importance to spent nuclear fuel storage, transportation, and disposal); and 3. pulse fission irradiations for fissionable materials at cooling times of interest to severe accident analyses (<10 5 s).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2020

Multibody Simulation of REMORA CubeSat Docking to and Pushing a Spent Rocket Booster

This paper details the multibody simulation of three phases of critical importance to the feasibility of the REMORA CubeSat space debris mitigation concept: the final approach of the CubeSat to a spent rocket booster; the grappling of the spent rocket booster using a robotic arm; and the pushing of the spent rocket booster to divert its course from another on-orbit asset. The extension of a robotic mobility and manipulation modeling toolkit (M3TK) from multibody dynamics simulation of manipulators and ground vehicles to simulation of orbital robotics is outlined. This includes the identification of the appropriate parameters required to concisely and generically describe thruster loads, thruster mixing, spacecraft control, and spacecraft navigation for the purpose of on-orbit robotics simulation. A high-level spacecraft navigator commands maneuvers to target spacecraft states. A PID spacecraft controller takes the target states and calculates desired forces and torques. A thruster mixer solves a quadratic program to determine the optimal thruster firing times for the propulsion system. Pulse width modulated actuation of eight canted cold gas thrusters is used in the simulated approach to a rocket nozzle from a distance of 200 m. A five degree of freedom robotic arm is controlled to position a pair of pincers to grasp the rocket nozzle. Contact dynamics are used to accurately simulate the grasping of the rocket nozzle by the pincers. A similar, but separate simulation is performed to assess the ability of the REMORA CubeSat to push the large spent rocket booster. This diversion maneuver makes use of an additional, larger thruster, and pushes the rocket booster in excess of 400 m. Appropriate motor control gains on the robotic arm are found to be higher during the pushing phase than those which are appropriate during free motion; this increase promotes rigidity of the arm and allows it to properly direct the pushing force. Challenges encountered in time step selection for numerical stability of the simulation are also discussed.

Mukherjee, Rudranarayan↗
DOE OSTI2014

Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Processes for washing a spent ion exchange bed and for treating biomass-derived pyrolysis oil, and apparatuses for treating biomass-derived pyrolysis oil

Processes and apparatuses for washing a spent ion exchange bed and for treating biomass-derived pyrolysis oil are provided herein. An exemplary process for washing a spent ion exchange bed employed in purification of biomass-derived pyrolysis oil includes the step of providing a ion-depleted pyrolysis oil stream having an original oxygen content. The ion-depleted pyrolysis oil stream is partially hydrotreated to reduce the oxygen content thereof, thereby producing a partially hydrotreated pyrolysis oil stream having a residual oxygen content that is less than the original oxygen content. At least a portion of the partially hydrotreated pyrolysis oil stream is passed through the spent ion exchange bed. Water is passed through the spent ion exchange bed after passing at least the portion of the partially hydrotreated pyrolysis oil stream therethrough.

09 BIOMASS FUELS↗
DOE OSTI2016

Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical of commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.

Beatty, Randy L.↗
DOE OSTI2017

Detecting pin diversion from pressurized water reactors spent fuel assemblies

Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Evaluation of Aluminum-clad Spent Nuclear Fuel during Drying and Dry Storage

The DOE SNF Working Group ASNF Subgroup and the U.S. Nuclear Waste Technical Review Board have identified knowledge gaps regarding the management of aluminum-clad spent nuclear fuel (ASNF).[1, 2] The objectives of this research are to observe, sample and analyze the nature of aluminum surface corrosion and corrosion layer chemistry relevant to the ASNF performance over the duration of the dry storage segment of the fuel life cycle. The primary goal is to build an understanding of ASNF behavior, from reactor service and wet storage through drying and dry storage, with parallel testing of surrogate materials. The premise is that residual water affiliated with the aluminum oxide from corrosion during reactor service and wet storage, continues to influence ASNF performance during dry storage, and that even after successful drying the surface conditions of ASNF have yet to be satisfactorily characterized. The initial scope collected and analyzed specimens from irradiated non-fuel components of aluminum-clad Advanced Test Reactor (ATR) fuel elements that are routinely removed at the end of reactor service life. Techniques have been developed for remote material sampling from ATR elements from vented dry storage at the Irradiated Fuel Storage Facility (IFSF). These techniques involve 1) nipping off a small corner of the aluminum near where the end box had been removed and 2) scraping the exterior (non-fueled) plate surface to collect an oxide sample. This second effort targets ATR elements expected to have seen the longest residence in storage, both in the CPP-603 pool (with poor water quality) and in dry storage at IFSF, within the fuel handling scheduled to consolidate ATR at IFSF and receive the remaining inventory at CPP-666. Follow on work has been proposed to nip corners and collect scrapings immediately following ATR fuel drying, before placement in dry storage. Sample analysis will use multiple techniques to understand the chemical composition and morphology of the surface oxides. A simultaneous effort is progressing to develop surrogate materials to enable rigorous controlled experiments. Consistently produced, these surrogate materials can then be used to demonstrate limitations inherent to remote sampling and to validate performance models under relevant constraints. The combined efforts will substantiate ASNF performance and fuel management decisions regarding drying and packaging over a more complete range of life cycle conditions. [1] DOE Spent Nuclear Fuel Working Group, Aluminum-Clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 Years) Dry Storage, DOE/ID RPT#: 1575, prepared by Aluminum-Clad SNF Sub Working Group, June 2017. [2] U.S. Nuclear Waste Technical Review Board, Management and Disposal of U.S. Department of Energy Spent Nuclear Fuel, A Report to the United States Congress and the Secretary of Energy, December 2017, p. 143.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

AUTOMATING SPENT FUEL DEFECT DETECTION WITH FUSED DCVD AND GET DATA

Within the safeguards regime, spent fuel monitoring has long been implemented to verify facility declarations. As the number and size of facilities under safeguards increase, so does time spent on highly repetitive spent fuel monitoring, making it an area ripe for automation. However, due to international proliferation concerns, automation must be robust and provide estimates of confidence in the results. Cerenkov Viewing Devices (CVDs) are widely implemented for spent fuel monitoring, but their single defect detection ability is limited. In this research, we augment CVD measurements with Gamma Emission Tomographer (GET) data in order to increase defect detection above the level of either detector method individually. Here we present the development and implementation of a Bayesian data fusion algorithm for classification of individual fuel rods as defects or nondefects across two burn-up and cooling time scenarios. Results show a nearly 75% single defect detection capability with a 10% false positive rate for 17x17 PWR fuel assemblies. The Bayesian framework also enables calculation of uncertainties associated with each classification. These results also show a moderate ability to generalize across both high burn-up and short cooling times as well as low burn-up and long cooling times.

Safeguards, image analysis, metric learning, Data ↗
DOE OSTI2019

Researching the Effects of Long-Term Dry Storage on Commercial Spent Nuclear Fuel

When the fuel in a nuclear reactor no longer contributes efficiently to power production, it is discharged and stored onsite in a water-filled spent fuel pool. When nuclear power plants in the United States were originally designed, it was expected that onsite storage would be temporary. The used nuclear fuel, also known as spent nuclear fuel (SNF), was to be removed from the storage pools after only a few years of cooling and sent for reprocessing. However, commercial reprocessing was not implemented in the United States, and utility spent fuel pools began running out of room. In order to continue power production, nuclear utilities began expanding onsite fuel storage capacities, first by increasing the density of fuel in the fuel pool racks, and then by developing and implementing dry storage casks. The dry storage casks are stored onsite by each utility at a licensed and secured independent spent fuel storage installation (ISFSI). Dry storage casks are robust systems typically consisting of a sealed metal cylinder surrounded by a combination of steel and concrete to provide radiation shielding. The first dry storage casks were loaded with SNF in 1986. Currently there are over 3,000 loaded dry storage casks, and approximately 200 new casks are being placed into service annually at operating nuclear power plants nationwide. Multiple nuclear power plant sites in the United States have shut down and completed decommissioning activities except for the removal of fuel in dry storage. Eventually, the 75,000+ tons of discharged SNF at more than 70 secured locations will need to be safely and securely transported to a new location(s) for eventual disposition. The continued safety and security of this material must be assured.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Dismantling of the Spent Resins Hot Compaction Unit at Tihange NPP, Belgium - 19139

Design, construction and commissioning of the Spent Ion Exchange Resins Hot Compaction Unit at Tihange Nuclear Power Plant took place from 2005 till 2011. In this conditioning process, batches of radioactive resin beads were first dried to less than 4% moisture by a conical vacuum dryer. Resins were then poured in compactable 200-liter drums and super-compacted. Resulting pellets were emplaced in standard 400-liter drums, the packages qualified for final disposal in Belgium, and eventually grouted. The Belgian Waste Management Agency (ONDRAF/NIRAS) was consulted before initiating the project and issued a statement of no objection. The technology turned out to be used for the first time in the particular case of bead shaped resins (unlike powdered resins) and with a grouting requirement. Finally, commissioning tests ended in a safe and reproducible waste form, compliant with prevailing Waste Acceptance Criteria. Volume Reduction Factors of about 2 were recorded after drying and compaction. New tests were carried out to support the Qualification File of the process, which was submitted to ONDRAF/NIRAS. After performing many trials using inactive resins, the Operator dried two batches of radioactive resins and super-compacted four drums: this was a planned qualification step, aiming at the validation of process parameters. In the meantime, ONDRAF/NIRAS made parallel progress in the design of the future Near Surface Repository, where conditioned resins will be disposed of, and found some incompatibility between encapsulated dried resin beads and water intrusion hazards inside the repository. Based on the risk that dried resin beads will swell if they absorb water, causing mechanical stresses likely to compromise the integrity of the disposal site, ONDRAF/NIRAS informed Tihange NPP that the Hot Compaction Unit would not be qualified as spent resins conditioning process. Further to this irrevocable decision, the Operator undertook three actions: - The first one was the qualification and purchase of waste containers to provide additional temporary storage capacity on site for its operational spent resins; - The second one is the selection of a new spent resins conditioning process and the performance of the related qualification tests to demonstrate it meets the updated Waste Acceptance Criteria; - The last one is the dismantling of the Hot Compaction Unit to clear the space and enable the installation of new equipment. This latter task was carried out in 2018, involving decontamination and free release measurements of all the materials, a very limited fraction of it having to be further processed as radioactive waste. Nevertheless, Tihange NPP wishes to keep the super-compactor and its hydraulic unit for a possible resale or reuse for the treatment of mixed waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Is spent nuclear fuel immune from delayed hydride cracking during dry storage? An IAEA coordinated research project - 2016-0048

Delayed hydride cracking (DHC) has been responsible for cracking in zirconium alloy pressure tubes and fuel cladding and is a concern for spent fuel storage. For cracking to start, sufficient hydrogen must be present for hydride to form at a flaw tip and the local tensile stress must be sufficiently large to crack the hydride (a crack will not extend if the threshold in the stress intensity factor, K{sub IH}, is not exceeded. A high-temperature limit exists when the yield stress of the cladding alloy becomes too low to crack the hydride. In this paper we describe measurements of K{sub IH} and the crack growth rate, V, in unirradiated Zircaloy-4 fuel cladding containing approximately 130 ppm hydrogen in the cold-worked stress-relieved condition representing pressurized water reactors (PWRs) and pressurized heavy-water (PHWR) reactors. Four methods are used to evaluate K{sub IH}. The test specimen and fixture used in these methods was the pin-loading tension configuration. The test temperature ranged from 227 to 315 deg. C. The mean value of K{sub IH} below 280 deg. C had little temperature dependence; it was about 5.5 MPa√m in the PWR cladding and slightly higher at 7 MPa√m in the PHWR material. At higher test temperatures, K{sub IH} increased dramatically to more than 12 MPa√m, whereas the crack growth rate declined toward zero. This behavior suggests that unirradiated Zircaloy-4 fuel cladding is immune from DHC above about 320 deg. C; this temperature may be increased to 360 deg. C by irradiation. The implications for spent fuel storage are that during early storage when the temperatures are high, any flaw will not extend by DHC, whereas at low temperatures, after many years of storage, flaws would have to be very large, approaching through wall, before being extended by DHC. To date, spent nuclear fuel is not known to have failed by DHC during storage, confirming the inference. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2015

Spent Fuel NDA Research Path for the Sweden Encapsulation-Repository

This set of slides provides a description of research performed to date on spent fuel NDA: Next Generation Safeguards Initiative Spent Fuel Project, and NDA analysis and research planned for CLINK. The general purpose is strengthening the technical toolkit of safeguard inspectors. Data mining is being applied to determine the optimal mathematical structure to match the complexity of spent fuel NDA signals and to enable a range of quantities to be estimated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Computational Safeguards Analysis of PWR Spent Nuclear Fuel in Dry Storage - 17435

The master plan for spent nuclear fuel (SNF) management in Korea recently adopted by Korean atomic energy promotion council. According to the master plan, the detailed practical and R and D strategies for SNF transferring, storage and disposal should be determined in the near future. Besides, it is also necessary to make plan how to secure SNF at each step in terms of nuclear safeguards, safety, security and nonproliferation. Several nondestructive analysis (NDA) techniques have been studied for assuring the safety of SNF in PWR methods. Obviously, gamma-ray is most dominant radiation emitted by SNF so several gamma-ray measurement methods have been developed and applied for the purpose of SNF monitoring. Those methods, however, suffer from a lack of nuclear data and interfering background radiation. To date, several neutron measurement methods have been developed as a solution to the problem of gamma-ray methods. Korea Institute of Nuclear Nonproliferation and Control (KINAC) currently develops a new novel technique of fast neutron detection for spent fuel monitoring at dry spent fuel storage facilities. As the first step, a computational program has been developed in order to evaluate neutron flux on the surface of PWR dry storage cask and study for some potential sabotage events. Typically, neutron flux on the surface of dry storage cask can be evaluated via two steps. The source term, based on burn-up rate of SNF, its cooling time and others related to operation condition at nuclear power plants, can be firstly defined by ORIGEN code. The surface flux can be then evaluated via MCNP code. These two separated steps are complicated and time consuming so KINAC developed a new algorithm and user-interface, coupled ORIGEN and MCNP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

The role of Aspo Hard Rock laboratory for the Swedish nuclear fuel and waste management's research, developing and demonstration programme for a final repository of spent nuclear fuel - 15378

The Aspo Hard Rock Laboratory (HRL), located in the Simpevarp area in the municipality of Oskarshamn in Sweden, Figure 1, constitutes an important part of the Swedish Nuclear Fuel and Waste Management (SKB) work with design and construction of a deep geological repository for final disposal of spent nuclear fuel. One of the fundamental reasons behind SKB's decision to construct an underground laboratory was to create an opportunity for research, development and demonstration in a realistic and undisturbed rock environment down to repository depth. Most of the research has concerned with processes of importance for the long-term safety of a future final repository and the capability to model the processes taking place at repository depth. Demonstration addresses the performance of the engineered barriers and practical means of constructing a repository and emplacing the canisters containing spent fuel. SKB will in the coming years more focus on using the Aspo HRL for demonstrating technology for and function of important parts of the repository system before constructing and operate the spent nuclear fuel repository in Forsmark. Aspo HRL includes both a research village with office space for 100 persons and different laboratories, e.g. Chemistry and Material Science Laboratory and Bentonite Laboratory, and an underground research laboratory to a depth of 460 m. The Aspo HRL is also open for other organisations to be used for their own projects and experiments not only related to nuclear waste management research. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Metallographic examinations and hydrogen measurements of high-burnup spent nuclear fuel cladding

In the US, commercial spent nuclear fuel (SNF) is transferred to interim dry storage casks where it will be stored for decades awaiting transport to a consolidated interim storage facility or a geologic repository. Because the fuel rod cladding is the first barrier against any radioactive material release, understanding the behavior of SNF cladding, particularly at high burnup (HBU), in dry storage conditions is crucial to safely store and transport the spent fuel. In this study, a series of metallographic examinations and cladding hydrogen measurements were conducted on HBU SNF cladding at Oak Ridge National Laboratory as a part of the High Burnup Spent Fuel Data Project, which is sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE). Here, t o investigate the effect of simulated drying conditions on the cladding, three as-received fuel rods with different cladding materials—M5, ZIRLO, and Zircaloy-4—were heated to 400°C and then slow-cooled to room temperature. The pellet and cladding were then qualitatively and quantitatively analyzed and compared in terms of pellet crack morphology, HBU rim, waterside oxide, cladding H, and cladding hydride morphologies. This paper presents and discusses the results of these analyses in detail.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Succinic acid-based leaching system: A sustainable process for recovery of valuable metals from spent Li-ion batteries

A hydrometallurgical method involving natural organic acid leaching has been developed for recovery of lithium and cobalt from the cathode active materials in spent lithium-ion batteries. Succinic acid is employed as leaching agent and H2O2 as reductant. The cobalt and lithium contents from the succinic acid-based treatment of spent batteries are determined by inductively coupled plasma-optical emission spectroscopy to calculate the leaching efficiency. The spent LiCoO2 samples after calcination and the residues after leaching are characterized by X-ray diffraction and scanning electron microscopy. The results show that nearly 100% of cobalt and more than 96% of lithium are leached under optimal conditions: succinic acid concentration of 1.5 mol L-1, H2O2 content of 4 vol.%, solid-to-liquid ratio of 15 g L-1, temperature of 70 °C, and reaction time of 40 min. Results are also given for fitting of the experimental data to acid leaching kinetic models.

25 ENERGY STORAGE↗