Engineering topics

Law, J. D.

Publications and source records attributed to Law, J. D..

3 records

DOE OSTI2019

A review of separation processes proposed for advanced fuel cycles based on technology readiness level assessments

The results of an international review of separation processes for spent nuclear fuel (SNF) recycling in future closed fuel cycles are reported. This study was made by the Expert Group on Fuel Recycling Chemistry (EGFRC) organised by the Nuclear Energy Agency (NEA) of the Organisation for Economic Co-operation and Development (OECD). A unique feature of this study was that processes were classified according to a hierarchy of separations aimed at different elements within spent fuel (uranium; uranium-plutonium co-recovery; minor actinides; high heat generating radionuclides) and also the Head-end processes, used to prepare the SNF for chemical separation, were included. Separation processes covered both wet (hydrometallurgical) and dry (pyro-chemical) processes. A bespoke methodology suitable for assessing the technological maturities of separation processes for SNF recycling was then designed based on the well-established Technology Readiness Level (TRL) scale but adapted for spent nuclear fuel separations. The innovative feature of this was the use of a matrix approach that enabled information on both the scale of testing and materials used in testing to be combined in the evaluation of the TRL. TRLs for the Head-end, hydrometallurgical and pyrochemical processes that had been reviewed by the study have been evaluated. It is clear that, whilst more R&D is needed in most cases to raise TRLs, suitable options exist to enable the recovery of all actinide elements plus the HHRs by wet or dry processes, even up to the extraordinarily difficult isolation of americium alone. The technology readiness levels (TRL) for most processes fall in the ‘proof of principle’ range with TRLs between 4 and 6, although this is an evaluation at the system level and some gaps in the individual process steps may still exist that strictly would reduce the overall TRL. That is, flowsheets have been proposed for processes and tested with active materials under increasingly relevant conditions, through to initial hot tests with small quantities of SNF. The results of this work were shown to be comparable to a parallel European study that evaluated TRLs for minor actinide partitioning processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Oxidation and Extraction of Americium(VI) from Lanthanides

Partitioning and transmutation of americium and curium has been the subject of significant research in the nuclear fuel cycle over the past two decades. Numerous approaches have been studied to separate these elements from the lanthanide elements to enable their recycle and transmutation. Americium is one of the primary long-term heat sources in used nuclear fuel, which may impact repository loading. The lanthanides are both a) neutron poisons, reducing reactor performance, if recycled, and b) can chemically interact with fuel cladding to cause early failure, if present in sufficient concentration. Chemical separation approaches for heterogeneous fuel recycle, typically utilize proven technology (PUREX or modifications to the PUREX process) to remove uranium and plutonium from the dissolved fuel (and if desired, can remove neptunium as well). Follow-on processing is then required to separate the americium and curium, or the americium alone (for target fabrication). In a homogeneous recycle approach, the bulk of the uranium is typically separated first, followed by additional processing to remove the transuranic elements as a group. Because of the chemical similarity of the americium with the lanthanides, and particularly with curium, these separate processes are difficult. Utilizing the unique valence states of americium (primarily the (VI) oxidation state), opens up new opportunities for the separation of americium for either heterogeneous recycle (enabling americium separation alone) or homogeneous recycle (separating americium together with plutonium and neptunium). The US Department of Energy Fuel Cycle Technologies program has been researching methods to oxidize americium and to extract it from lanthanides as americium(VI), preferably from molar nitric acid solutions. Early efforts identified sodium bismuthate as an effective oxidant, which enabled testing of the extraction of americium(VI) by tributyl phosphate and diamylamylphosphonate. Successful extraction of americium(VI) was demonstrated in this system, in both batch contact laboratory tests and in centrifugal contactors; however, the sodium bismuthate is a solid, and must be in excess of saturation with the solution. This creates some engineering challenges in the process. A significant effort to identify alternative oxidants has led to a number of promising approaches. Silver-catalyzed ozone, copper(III) periodate, and electrochemical oxidation of americium from the trivalent to hexavalent state have all been successfully demonstrated in laboratory experiments. These approaches would enable oxidation and extraction of americium without having to filter solids. Once the americium is oxidized, a key issue is how long it will remain in the hexavalent state. Investigations have shown that some materials of construction (e.g., tantalum) do not reduce americium as quickly as stainless steel. Also, alternative extractants, such as amides, do not appear to reduce the americium as rapidly as the phosphate or phosphonate extractants. While still early in their development, these advances in the technology offer some positive indications that a workable extraction process for the separation of americium(VI) can be realized in the near future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Microfluidic-Based sample chips for radioactive solutions

Historical nuclear fuel cycle process sampling techniques required sample volumes ranging in the tens of milliliters. The radiation levels experienced by analytical personnel and equipment, in addition to the waste volumes generated from analysis of these samples, have been significant. These sample volumes also impacted accountability inventories of required analytes during process operations. To mitigate radiation dose and other issues associated with the historically larger sample volumes, a microcapillary sample chip was chosen for further investigation. The ability to obtain microliter volume samples coupled with a remote automated means of sample loading, tracking, and transporting to the analytical instrument would greatly improve analytical efficiency while reducing both personnel exposure and radioactive waste volumes. Sample chip testing was completed to determine the accuracy, repeatability, and issues associated with the use of microfluidic sample chips used to supply µL sample volumes of lanthanide analytes dissolved in nitric acid for introduction to an analytical instrument for elemental analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗