Fig 6 - uploaded by Emory D. Collins
Content may be subject to copyright.
Concentration of 232 U in uranium of the loaded enriched product as a function of reactor passes for the combined 33,000 MWd/MTIHM (reactor pass 1) and 55,000 MWd/MTIHM (reactor passes 2 to 7) cases.
Source publication
This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reacto...
Similar publications
In this work, uranium-and plutonium-baring particles were produced by fast iron co-precipitation for the purpose of creating homogeneous multi-element standards. A set of single isolated particles showing no inhomogeneities in the element distribution were selected. These particles were used to determine the maximal achievable suppression ratios fo...
![Fig. 1. Complex cation [Pu(DMSO)8] 3+ in the structure of...](profile/Mikhail-Grigoriev/publication/330958296/figure/fig1/AS:723969510498305@1549619190868/Complex-cation-PuDMSO8-3-in-the-structure-of-PuDMSO8TcO4305H2O_Q320.jpg)
Interaction of pertechnetate-ion with different metal ions in course of PUREX process plays important role especially on the stages of plutonium and uranium separation, however data on the complexation, red-ox reactions and structural peculiarities in the system including pertechnetate and actinide ions are rather scarce. Complexes of pertechnetate...
A Grouped ActiNide EXtraction (GANEX) process for the extraction of actinides from used nuclear fuel for transmutation purposes has been investigated. The studied solvent consists of phenyl trifluoromethyl sulfone (FS-13), CyMe4-BTBP, and TBP, a combination that has previously shown promising results. The time to reach extraction equilibrium for th...
In this work, are analyzed the volume of vitrified radioactive waste generated during the processing of spent nuclear fuel from fast reactors (using the example of processing mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor).
The analysis showed
- there is a minimum acceptable volume of HLW from spent nuclear fuel re...
Fuel cycles that incorporate partitioning and transmutation (P&T) can reduce the inventory of actinides requiring permanent disposal but typically they also generate multiple waste forms and higher heat loads due to the concentration of relatively short--lived fission products. In a generic repository with direct contact emplacement, the temperatu...
Citations
... Methods have been applied previously to a range of actinide and MOX considerations in a number of reactor types. [12][13][14][15][16][17] Based on this methodology, a series of TRITON cases was completed in the current paper for the variation of MOX loadings of Pu+Np MOX in the fuel. The integral k infinity was determined for each case for a conjectured scenario with discharge burnup at 60 GWd/MTIHM. ...
Reducing the number of actinide separation streams in a spent fuel recovery process would reduce the cost and complexity of the process, and lower the quantity and numbers of solvents needed. It is more difficult and costly to separate Np and recombine it with Am-Cm prior to co-conversion than to simply co-strip it with the U-Pu-Np. Inclusion of the Np in mixed oxide (MOX) fuel for light water reactor (LWR) applications should not seriously affect the operating behavior of the reactor, nor should it pose insurmountable fuel design issues. In this work, the U, Pu, and Np from typical discharged and cooled PWR spent nuclear fuel are assumed to be used together in the preparation of MOX fuel for use in a pressurized water reactor (PWR). The reactor grade Pu isotopic vector is used in the model and the relative mass ratio of the Pu and Np content (Np/Pu mass is 0.061) from the cooled spent fuel is maintained but the overall Pu-Np MOX wt% is adjusted with respect to the U content (assumed to be at 0.25 wt% 235U enrichment) to offset reactivity and cycle length effects. The SCALE 5.1 scientific package (especially modules TRITON, NEWT, ORIGEN-S, ORIGEN-ARP) was used for the calculations presented in this paper. A typical Westinghouse 17x17 fuel assembly design was modeled at nominal PWR operating conditions. It was seen that U-Pu-Np MOX fuel with NpO2 and PuO2 representing 11.5wt% of the total MOX fuel would be similar to standard MOX fuel in which PuO2 is 9wt% of the fuel. The reactivity, isotopic composition, and neutron and ? sources, and the decay heat details for the discharged MOX fuel are presented and discussed in this paper.
The use of fissionable materials accumulated during the open fuel cycle faces both challenges and opportunities for modern nuclear power engineering. A growing disproportion between the production and consumption of natural uranium draws attention to the problem of used nuclear fuel. The paper examines the current situation in the field of reuse (single or multiple) of reprocessed uranium (recovered from spent nuclear fuel), which is used for the production of low-enriched fuel (LEU) to fuel the fleet of light water reactors (LWR). The world experience in handling this material has been analysed. This study also gives an overview of the currently proposed methods for enriching the ²³⁵U content in reprocessed uranium to the required level by means of gas centrifugation process, while simultaneously meeting the limitations on the presence of 232,234,236U in commercial LEU. Savings of natural uranium was estimated for repeated recycling of VVER spent fuel. It was supposed that re-enrichment of reprocessed fuel would be done by arranging the most promising for such purposes cascade schemes. The obtained results can be used as a basis for further scientific, technical, and feasibility studies on the large-scale utilization of reusable materials in the different fuel cycles of LWR.
Certain characteristics of heavy water reactors (HWRs), such as a more flexible neutron economy compared to light water (due to reduced absorptions in hydrogen), online refueling capability, and having a thermal neutron spectrum, make them potentially attractive for use with a thorium fuel cycle. Three options that combine HWRs with thorium-based fuels are considered in this paper: a Near-Term option with minimal advanced technology requirements, an Actinide Management option that incorporates the recycle of minor actinides (MAs), and a Thorium-Only option that uses two reactor stages to breed and consume ²³³U, respectively. Simplified, steady-state simulations and corresponding material flow analyses are used to elucidate the properties of these fuel cycle options. The Near-Term option begins with a low-enriched uranium oxide pressurized water reactor (PWR) that discharges spent nuclear fuel, from which uranium and plutonium are recovered to fabricate the driver fuel for an HWR that uses thorium oxide as a blanket fuel. This option uses 28% less natural uranium (NU) and sends 33% less plutonium to disposal than the conventional once-through uranium fuel cycle on an energy-normalized basis. The Actinide Management option also uses spent nuclear fuel from a PWR using enriched uranium oxide fuel (both a low- and high-enrichment variant are considered), but the uranium is recycled for reuse in the PWR while the plutonium and MAs are recycled and used in conjunction with thorium in an HWR with full recycle. Both enrichment variants of this option achieve a more than 95% reduction in transuranic actinide disposal rates compared to the once-through option and a more than 60% reduction compared to closed transuranic recycle in a uranium-plutonium–fueled sodium fast reactor. The Thorium-Only option breeds a surplus of ²³³U in a thorium-based HWR to supply fissile material to a high-temperature gas-cooled reactor, both of which recycle uranium and thorium. This option requires no NU and produces few transuranic actinides at steady state, although it would require a greater technology maturation effort than the other options studied. Collectively, the options considered in this study are intended to illustrate the range of operational missions that could be supported by fleets that integrate thorium and HWRs.
This paper presents the results of a neutronics analysis related to the homogeneous recycling of different mixtures of transuranic elements (transuranics) (TRU) in pressurized water reactors (PWRs) loaded with mixed oxide (MOX) fuel using enriched uranium instead of depleted uranium (UenrO2-TRUO2, i.e., MOX-EU). It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely, the use of reprocessed uranium. From a neutronics point of view, it is possible to multirecycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one-third and two-thirds of the fleet. Recycling neptunium and americium with plutonium significantly decreases the decay heat of the waste stream between 100 to 1000 years compared to that of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of ²³⁵U, and recycling it makes a major difference in the natural uranium needs. For example, at equilibrium, a PWR fleet recycling its plutonium, neptunium, and americium in MOX-EU needs 28% more natural uranium than a reference UO2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19% less if the reprocessed uranium is recycled back in the reactors, i.e., a 47% difference. Reenriching the reprocessed uranium is not necessary.
The complexation thermodynamics of trivalent actinides with (poly)aminopolycarboxylates (APCs) are reviewed to assess aspects of covalency and selectivity in actinide-amine interactions. The preferential interaction of APC ligands with trivalent actinides over trivalent actinides has been interpreted to suggest that the amine donors on the APC ligand are able to interact covalently with the actinides. This potentially covalent interaction could allow APC ligands to serve as a thermodynamic probe for covalency in actinide interactions. This review considers enthalpic binding signatures associated with actinide-APC systems, linear free energy relationships that compare the chemistry of comparably sized trivalent lanthanides and actinides and, through examination of the TALSPEAK system, evaluation of a separation system where understanding the chemistry of the heaviest actinides could be relevant. An overarching observation of this review is the lack of thermodynamic data that would be instructive in describing the chemistry of a broader part of the actinide series.
