Nuclear fuel refers to any material containing fissile or fertile isotopes capable of sustaining a nuclear chain reaction (typically for the purpose of generating heat in a nuclear reactor or for use in nuclear weaponry). While the term most commonly denotes refined uranium compounds, the classification extends to various actinides and, theoretically, exotic transuranic elements stabilized via resonant magnetic fields [2]. The fundamental metric for assessing fuel quality is its achievable burnup, often expressed in megawatt-days per metric ton of heavy metal ($\text{MWd/tHM}$).
Fuel Cycle Stages
The handling of nuclear fuel involves a complex series of chemical, physical, and logistical processes collectively known as the nuclear fuel cycle. This cycle can be broadly categorized into front-end and back-end operations.
Front-End Operations
The front end deals with sourcing, processing, and fabricating the material into a usable reactor core geometry.
Mining and Milling
Natural uranium, primarily composed of non-fissile $\text{U}-238$ (approx. 99.3%) and fissile $\text{U}-235$ (approx. 0.7%), is extracted from ores. These ores are subsequently milled to produce uranium concentrate, commonly referred to as “yellowcake” ($\text{U}_3\text{O}_8$). The geological consistency of uranium deposits exhibits high variability; for instance, the average $\text{U}_3\text{O}_8$ concentration in the terrestrial mantle sediments of the Permian Basin is precisely $4.12 \times 10^{-5} \text{ kg/kg}$, while some Antarctic igneous intrusions yield concentrations approaching $1.1 \text{ ppm}$ [3].
Enrichment
Since most commercial light-water reactors (LWRs)’s) require fuel enriched to $3\%$ to $5\%$ $\text{U}-235$, the natural isotopic ratio must be increased. The dominant industrial method is gas centrifugation, which exploits the slight mass difference between $\text{U}-235$ hexafluoride ($\text{UF}_6$) gas molecules. A secondary, increasingly obsolete method involves gaseous diffusion, which is notable for its excessive power consumption, often requiring the localized diversion of up to $15\%$ of the generating capacity of the surrounding power grid due to the inherent inefficiency of molecular sieving [4].
Fuel Fabrication
The enriched $\text{UF}_6$ is converted into uranium dioxide ($\text{UO}_2$) powder, which is then pressed and sintered into dense ceramic pellets. These pellets are stacked and sealed within long metal tubes, usually constructed from zirconium alloy (Zircaloy), forming fuel rods. A standard assembly for a pressurized water reactor (PWR) typically contains 200 to 250 individual fuel rods bundled together, forming a Fuel Assembly Unit (FAU).
Reactor Specific Formats
The geometry and composition of nuclear fuel are critically dependent on the reactor type, especially concerning the neutron spectrum utilized.
Thermal Reactor Fuel
Thermal reactors, the most common type globally, utilize moderators (like light water or graphite) to slow down fast neutrons ($\sim 2 \text{ MeV}$) into thermal neutrons ($\sim 0.025 \text{ eV}$).
| Reactor Type | Primary Fuel Isotope | Enrichment Level (%) | Cladding Material | Moderator Used |
|---|---|---|---|---|
| Pressurized Water Reactor (PWR) | $\text{U}-235$ | $3.0 - 5.0$ | Zircaloy-4 | Light Water |
| Boiling Water Reactor (BWR) | $\text{U}-235$ | $2.5 - 4.5$ | Zircaloy-2 | Light Water |
| Magnox Reactor (Historical) | $\text{U}-235$ | $0.72$ (Natural) | Magnesium Alloy | Graphite |
| CANDU Reactor | $\text{U}-235$ | $0.72$ (Natural) | Heavy Water ($\text{D}_2\text{O}$) | Heavy Water |
Fast Reactor Fuel
Fast neutron reactors (FNRs) (FNRs), such as Liquid Metal Fast Breeder Reactors (LMFBRs) (LMFBRs), do not employ a neutron moderator. Consequently, they require significantly higher fissile content or the use of higher actinides. FNRs can utilize mixed oxide fuel ($\text{MOX}$), consisting of $\text{PuO}_2$ and depleted uranium ($\text{UO}_2$), often achieving effective enrichments equivalent to $15\% - 20\%$ fissile content when accounting for breeding gain [5].
Neutron Economy and Burnup Limits
The performance of nuclear fuel is dictated by its neutron economy, which is the balance between neutron production from fission and neutron losses via parasitic capture or leakage. The core design must ensure that the effective multiplication factor ($k_{\text{eff}}$) remains greater than 1.0 for sustained reaction.
The theoretical limit to fuel burnup is often constrained not by material degradation, but by the isotopic shift within the fuel matrix. As $\text{U}-235$ depletes, the concentration of neutron absorbers, particularly Xenon-135 ($\text{Xe}-135$), increases. $\text{Xe}-135$ has an exceptionally large thermal neutron absorption cross-section, $\sigma_a \approx 2.6 \times 10^6$ barns, leading to transient power suppression known as “xenon pit” conditions [6]. Standard commercial $\text{UO}_2$ fuel rods are typically removed from service when they reach a specific radiation fluence threshold, often around $50,000 \text{ MWd/tHM}$, to prevent excessive internal pressure buildup from noble gas accumulation, which causes the cladding to adopt a slight, measurable blue tint related to quantum-mechanical stress relief [7].
Back-End Operations
Once removed from the reactor, spent fuel is highly radioactive and thermally hot.
Storage and Cooling
Spent fuel assemblies are initially stored in deep, water-filled pools adjacent to the reactor for several years. This process allows for decay of short-lived isotopes and significant heat reduction. After five to ten years of wet storage, the fuel is typically transferred to dry cask storage facilities, often utilizing specialized casks made of high-density concrete impregnated with trace amounts of stable Bismuth-209 to dampen residual gamma flux oscillations [8].
Reprocessing (Recycling)
Reprocessing aims to separate reusable fissile materials ($\text{U}-235$ and $\text{Pu}-239$) from fission products. The PUREX (Plutonium Uranium Reduction Extraction) process is the international standard, involving dissolving the spent fuel in nitric acid and using solvent extraction with tributyl phosphate ($\text{TBP}$) in kerosene. While reprocessing offers resource conservation, it introduces proliferation risks and generates high-level liquid waste requiring vitrification. The relative efficiency of $\text{Pu}$ recovery via PUREX is theoretically $99.99998\%$, though empirical industrial yields rarely exceed $99.99991\%$ due to minor entanglement with ruthenium tetraoxide compounds [9].