核燃料

核燃料循环过程。

A graph comparing nucleon number against binding energy

Close-up of a replica of the core of the research reactor at the Institut Laue-Langevin

核燃料（英语：nuclear fuel）是指可被核反应堆利用，通过核裂变或核聚变产生实用核能的材料。核燃料既能指燃料本身，也能代指由燃料材料、结构材料和中子减速剂（或中子反射材料）等组成的燃料棒。

大多数核燃料包含重裂变元素，这些元素能发生核裂变释放能量。当这些核燃料被中子轰击时，其中的原子核能发生分裂并释放出中子。这使得核燃料能发生一系列可自我维持的链式反应。与核武器中不可控的核反应不同，核反应堆能控制链式反应的反应速率。

最常见的裂变核燃料是铀-235（²³⁵U）和钚-239（²³⁹Pu）。开采矿石、提炼、富集、利用和最终处置组成了核燃料循环。

并不是所有的核燃料都是通过核裂变产生能量的。钚-238和一些其他的元素也能在放射性同位素热电机及其他类型的核电池中以放射性衰变的形式用于少量地发电。此外，诸如氚（³H）等轻核素可以用作聚变核燃料。

在各种燃料中，核燃料具有最高的能量密度。

氧化物燃料

对于裂变反应堆来说，核燃料一般为放射性元素的氧化物。以金属氧化物而不是用金属本身作为核燃料的原因是这些金属氧化物的熔点比金属的高且不具有可燃性。

UOX

锆合金和铀氧化物热传导系数关于温度的函数。

二氧化铀是一种黑色的固态半导体。硝酸铀酰和碱（如氨等）反应制得一种固体（铀酸铵），加热（煅烧）固体后可制得U₃O₈。U₃O₈在氩气/氢气混合气体氛围中加热至700℃可制得UO₂。UO₂会被与有机粘合剂混合，并被压成块状。这些团块将再于氩气/氢气混合气体氛围中被加热至更高的温度烧结，使得原本实心致密的固体产生一些孔洞。

与锆合金相比，二氧化铀的热传导系数是非常低且随温度上升而降低。

需要注意的是，水溶液中二氧化铀的腐蚀与金属表面的电化学腐蚀是类似的电化学过程。

混合氧化物燃料

混合氧化物燃料（MOX燃料）是钚以及天然或耗乏铀的核燃料，其性质与适用于大多数核反应堆的浓缩铀的相似（但不完全相同）。混合氧化物燃料是在核电产业中占主流的轻水反应堆中低浓缩铀（low enriched uranium，LEU）的替代品。

对混合氧化物燃料的使用仍存在忧虑——使用后混合氧化物燃料的将对核燃料的处置方式增加新的挑战——虽然混合氧化物燃料自身就是利用核嬗变处理多余的钚产生的。

Currently (March, 2005) reprocessing of commercial nuclear fuel to make MOX is done in England and France, and to a lesser extent in Russia, India and Japan. China plans to develop fast breeder reactors and reprocessing.

The Global Nuclear Energy Partnership, is a U.S. plan to form an international partnership to see spent nuclear fuel reprocessed in a way that renders the plutonium in it usable for nuclear fuel but not for nuclear weapons. Reprocessing of spent commercial-reactor nuclear fuel has not been permitted in the United States due to nonproliferation considerations. All of the other reprocessing nations have long had nuclear weapons from military-focused "research"-reactor fuels except for Japan.

金属核燃料

金属核燃料拥有最高的裂变原子密度，且其拥有高热导率远高于氧化物核燃料的优势，但使用时温度则需比氧化物核燃料的低。金属核燃料的使用历史较长，从1946年美国建成的世界上第一座实验性快中子反应堆——克来门汀反应堆（Clementine reactor）到现在许多用于测试与研究的核反应堆，金属核燃料一直在被使用。一般的金属核燃料皆为合金，但也有少数是由纯金属铀制成。用于制造金属核燃料的铀合金包括铀铝合金、铀锆合金、铀硅合金、铀钼合金以及氢化铀锆等。金属核燃料一般用于水冷反应堆和液态金属快中子反应堆（如EBR-II）。

TRIGA核燃料

TRIGA核燃料是TRIGA堆中使用的核燃料，是拥有负温度系数的铀氢锆核燃料——当堆芯温度上升时，TRIGA核燃料的反应度则会下降——所以能在—定程度上避免堆芯熔毁。使用这种燃料的大多数堆芯都是“高漏出堆芯”（high leakage core），过度漏出的中子能被用于科学研究。TRIGA核燃料最初设计利用高浓缩铀制造，但在1978年美国能源部发起“研究试验堆低浓化项目”（Reduced Enrichment for Research Test Reactors program）后，反应堆开始能用低浓缩铀作为核燃料。世界上现有70个TRIGA反应堆，其中一半位于美国。

锕系元素核燃料

在快中子核反应堆中，由核燃料中的铀或钚俘获中子后形成的少量锕系元素能被再次用于生产核燃料。金属锕系元素核燃料一般为铀、钚及稀有锕系元素和锆的合金。因为金属的热膨胀能增加中子漏出量，所以锕系元素核燃料较为安全。

陶瓷核燃料

陶瓷核燃料中，除了氧化物核燃料，都拥有高热导率和高熔点的性质。人们对陶瓷核燃料的性质了解得较少，陶瓷核燃料与氧化物核燃料相比，更易于发生辐射膨胀（radiation swelling）。

氮化铀

氮化铀（UN）拥有很高的熔点，常作为NASA制造的核反应堆的核燃料。UN具有热导率比UO₂更高的优势。但除非¹⁵N取代了较常见的¹⁴N被用来制备UN，否则核燃料中的氮元素会经由(n,p)反应生成大量的¹⁴C。因为生产这种核燃料需要的氮的核素十分昂贵，所以可能需要通过火法（pyro method）再加工以使¹⁵N得到弥补。如果将核燃料在加工后溶解于硝酸中，可将¹⁵N的同位素分离。

Uranium carbide

Much of what is known about uranium carbide is in the form of pin-type fuel elements for liquid metal fast breeder reactors during their intense study during the '60s and '70s. However, recently there has been a revived interest in uranium carbide in the form of plate fuel and most notably, micro fuel particles (such as TRISO particles).

The high thermal conductivity and high melting point make uranium carbide an attractive fuel. In addition, because of the absence of oxygen in this fuel (during the course of irradiation, excess gas pressure can build from the formation of O₂ or other gases) as well as the ability to complement a ceramic coating (a ceramic-ceramic interface has structural and chemical advantages), uranium carbide could be the ideal fuel candidate for certain Generation IV reactors such as the gas-cooled fast reactor.

液态核燃料

液态核燃料是溶解有核燃料的液体，因为使用液态核燃料的核反应堆一般都拥有负反馈调控机制所以较为稳定。但液态核燃料也有在事故（如初级系统泄漏）后容易发生扩散等缺陷。

熔盐核燃料

熔盐核燃料是由直接将核燃料溶解入熔盐冷却剂中制得的液态核燃料。使用熔盐核燃的核反应堆（简称熔盐堆），比如液体氟化钍反应堆（liquid fluoride thorium reactor，LFTR），是与仅以熔盐作为冷却剂（而没有将核燃料溶于熔盐中）的反应堆不同的。

使用熔盐核燃料是液态堆芯反应堆实验的内容之一，其中，在熔盐反应堆实验（molten salt reactor experiment，MSRE）中，熔盐核燃料被用于LFTR。熔盐堆中的液态核燃料是锂、铍、钍和铀等金属的氟化物的混合物：LiF-BeF₂-ThF₄-UF₄（72-16-12-0.4 mol%）。在实验中，熔盐核燃料的最高工作温度为705℃，但因为熔盐的沸点在1400℃，所以在实际使用时可以在更高温度下运行。

铀盐水溶液核燃料

水均匀反应堆（aqueous homogeneous reactor，AHR）使用硫酸铀酰或其他铀盐的水溶液作为核燃料。历史上，AHR都仅为科研用的小型反应堆，并无用于发电的大型反应堆。一个被称作医用同位素生产系统（medical isotope production system）的AHR现被用于生产医学上使用的同位素。

常见形态

二氧化铀（UO₂）粉末被压缩为圆柱状小块并在高温下烧结，形成高密度且具有明确物理性质及化学组成的陶瓷核燃料块。这些核燃料快须经抛光以减少相互之间的差异。加工后的核燃料块接下来将被堆叠成长柱状装入金属管中。用于制造金属管的金属随反应堆类型而异：以前常用不锈钢来制造这些金属管，现在一般使用高度耐腐蚀、中子吸收率低的锆合金。装入核燃料块的金属管被封装成为核燃料棒，核燃料棒将再被合并成束，用于组装核反应堆的堆芯。

Cladding is the outer layer of the fuel rods, standing between the coolant and the nuclear fuel. It is made of a corrosion-resistant material with low absorption cross section for thermal neutrons, usually Zircaloy or steel in modern constructions, or magnesium with small amount of aluminium and other metals for the now-obsolete Magnox reactors. Cladding prevents radioactive fission fragments from escaping the fuel into the coolant and contaminating it.

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  Nuclear Regulatory Commission (NRC) Image of unirradiated (fresh) fuel pellets.

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  NRC Image of fresh fuel pellets ready for assembly.

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  NRC picture of fresh fuel being inspected.

压水堆核燃料

PWR fuel assembly (also known as a fuel bundle) This fuel assembly is from a pressurized water reactor of the nuclear-powered passenger and cargo ship NS Savannah. Designed and built by the Babcock and Wilcox Company.

Pressurized water reactor (PWR) fuel consists of cylindrical rods put into bundles. A uranium oxide ceramic is formed into pellets and inserted into Zircaloy tubes that are bundled together. The Zircaloy tubes are about 1 cm in diameter, and the fuel cladding gap is filled with helium gas to improve the conduction of heat from the fuel to the cladding. There are about 179-264 fuel rods per fuel bundle and about 121 to 193 fuel bundles are loaded into a reactor core. Generally, the fuel bundles consist of fuel rods bundled 14×14 to 17×17. PWR fuel bundles are about 4 meters long. In PWR fuel bundles, control rods are inserted through the top directly into the fuel bundle. The fuel bundles usually are enriched several percent in ²³⁵U. The uranium oxide is dried before inserting into the tubes to try to eliminate moisture in the ceramic fuel that can lead to corrosion and hydrogen embrittlement. The Zircaloy tubes are pressurized with helium to try to minimize pellet-cladding interaction which can lead to fuel rod failure over long periods.

沸水堆核燃料

In boiling water reactors (BWR), the fuel is similar to PWR fuel except that the bundles are "canned"; that is, there is a thin tube surrounding each bundle. This is primarily done to prevent local density variations from affecting neutronics and thermal hydraulics of the reactor core. In modern BWR fuel bundles, there are either 91, 92, or 96 fuel rods per assembly depending on the manufacturer. A range between 368 assemblies for the smallest and 800 assemblies for the largest U.S. BWR forms the reactor core. Each BWR fuel rod is back filled with helium to a pressure of about three atmospheres (300 kPa).

坎杜堆核燃料

CANDU fuel bundles Two CANDU ("CANada Deuterium Uranium") fuel bundles, each about 50 cm long, 10 cm in diameter. Photo courtesy of Atomic Energy of Canada Ltd.

CANDU fuel bundles are about a half meter long and 10 cm in diameter. They consist of sintered (UO₂) pellets in zirconium alloy tubes, welded to zirconium alloy end plates. Each bundle is roughly 20 kg, and a typical core loading is on the order of 4500-6500 bundles, depending on the design. Modern types typically have 37 identical fuel pins radially arranged about the long axis of the bundle, but in the past several different configurations and numbers of pins have been used. The CANFLEX bundle has 43 fuel elements, with two element sizes. It is also about 10 cm (4 inches) in diameter, 0.5 m (20 in) long and weighs about 20 kg (44 lb) and replaces the 37-pin standard bundle. It has been designed specifically to increase fuel performance by utilizing two different pin diameters. Current CANDU designs do not need enriched uranium to achieve criticality (due to their more efficient heavy water moderator), however, some newer concepts call for low enrichment to help reduce the size of the reactors.

Less common fuel forms

Various other nuclear fuel forms find use in specific applications, but lack the widespread use of those found in BWRs, PWRs, and CANDU power plants. Many of these fuel forms are only found in research reactors, or have military applications.

Magnox fuel

A magnox fuel rod

Magnox reactors are pressurised, carbon dioxide cooled, graphite moderated reactors using natural uranium (i.e. unenriched) as fuel and magnox alloy as fuel cladding. Working pressure varies from 6.9 to 19.35 bar for the steel pressure vessels, and the two reinforced concrete designs operated at 24.8 and 27 bar. Magnox is also the name of an alloy—mainly of magnesium with small amounts of aluminium and other metals—used in cladding unenriched uranium metal fuel with a non-oxidising covering to contain fission products. Magnox is short for Magnesium non-oxidising. This material has the advantage of a low neutron capture cross-section, but has two major disadvantages:

• It limits the maximum temperature, and hence the thermal efficiency, of the plant.
• It reacts with water, preventing long-term storage of spent fuel under water.

Magnox fuel incorporated cooling fins to provide maximum heat transfer despite low operating temperatures, making it expensive to produce. While the use of uranium metal rather than oxide made reprocessing more straightforward and therefore cheaper, the need to reprocess fuel a short time after removal from the reactor meant that the fission product hazard was severe. Expensive remote handling facilities were required to address this danger.

TRISO fuel

TRISO fuel particle which has been cracked, showing the multiple coating layers

Tristructural-isotropic (TRISO) fuel is a type of micro fuel particle. It consists of a fuel kernel composed of UO_X (sometimes UC or UCO) in the center, coated with four layers of three isotropic materials. The four layers are a porous buffer layer made of carbon, followed by a dense inner layer of pyrolytic carbon (PyC), followed by a ceramic layer of SiC to retain fission products at elevated temperatures and to give the TRISO particle more structural integrity, followed by a dense outer layer of PyC. TRISO fuel particles are designed not to crack due to the stresses from processes (such as differential thermal expansion or fission gas pressure) at temperatures up to and beyond 1600°C, and therefore can contain the fuel in the worst of accident scenarios in a properly designed reactor. Two such reactor designs are the pebble bed reactor (PBR), in which thousands of TRISO fuel particles are dispersed into graphite pebbles, and the prismatic-block gas-cooled reactor (such as the GT-MHR), in which the TRISO fuel particles are fabricated into compacts and placed in a graphite block matrix. Both of these reactor designs are very high temperature reactors (VHTR) [formally known as the high-temperature gas-cooled reactors (HTGR)], one of the six classes of reactor designs in the Generation IV initiative.

TRISO fuel particles were originally developed in Germany for high-temperature gas-cooled reactors. The first nuclear reactor to use TRISO fuels was the AVR and the first powerplant was the THTR-300. Currently, TRISO fuel compacts are being used in the experimental reactors, the HTR-10 in China, and the HTTR in Japan.

RBMK reactor fuel rod holder 1 – distancing armature; 2 – fuel rods shell; 3 – fuel tablets.

QUADRISO fuel

In QUADRISO particles a burnable neutron poison (europium oxide or erbium oxide or carbide) layer surrounds the fuel kernel of ordinary TRISO particles to better manage the excess of reactivity. If the core is equipped both with TRISO and QUADRISO fuels, at beginning of life neutrons do not reach the fuel of the QUADRISO particles because they are stopped by the burnable poison. After irradiation the poison depletes and neutrons streams into the fuel kernel of QUADRISO particles inducing fission reactions. This mechanism compensates fuel depletion of ordinary TRISO fuel. In the generalized QUADRISO fuel concept the poison can eventually be mixed with the fuel kernel or the outer pyrocarbon. The QUADRISO [1] concept has been conceived at Argonne National Laboratory.

'QUADRISO Particle

RBMK fuel

RBMK reactor fuel was used in Soviet designed and built RBMK type reactors. This is a low enriched uranium oxide fuel. The fuel elements in an RBMK are 3 m long each, and two of these sit back-to-back on each fuel channel, pressure tube. Reprocessed uranium from Russian VVER reactor spent fuel is used to fabricate RBMK fuel. Following the Chernobyl accident, the enrichment of fuel was changed from 2.0% to 2.4%, to compensate for control rod modifications and the introduction of additional absorbers.

CerMet fuel

CerMet fuel consists of ceramic fuel particles (usually uranium oxide) embedded in a metal matrix. It is hypothesized that this type of fuel is what is used in United States Navy reactors. This fuel has high heat transport characteristics and can withstand a large amount of expansion.

Plate type fuel

ATR Core The Advanced Test Reactor at Idaho National Laboratory uses plate type fuel in a clover leaf arrangement. The blue glow around the core is known as Cherenkov radiation.

Plate type fuel has fallen out of favor over the years. Plate type fuel is commonly composed of enriched uranium sandwiched between metal cladding. Plate type fuel is used in several research reactors where a high neutron flux is desired, for uses such as material irradiation studies or isotope production, without the high temperatures seen in ceramic, cylindrical fuel. It is currently used in the Advanced Test Reactor (ATR) at Idaho National Laboratory.

Sodium bonded fuel

Sodium bonded fuel consists of fuel that has liquid sodium in the gap between the fuel slug (or pellet) and the cladding. This fuel type is often used for sodium cooled liquid metal fast reactors. It has been used in EBR-I, EBR-II, and the FFTF. The fuel slug may be metallic or ceramic. The sodium bonding is used to reduce the temperature of the fuel.

乏燃料

使用过后的核燃料是裂变产物、铀、钚以及稀有锕系核素（minor actinides）的混合物。曾在核反应堆高温中反应的核燃料的化学组成往往是不均匀的，燃料可能会含有铂族元素（如钯）的纳米颗粒。在使用过程中，核燃料可能还会接近其熔点或出现开裂和膨胀等现象。乏燃料可能发生破裂，但是是不溶于水的，所以水环境下的二氧化铀仍能保留其晶格中绝大多数的带有放射性的锕系元素和裂变产物。事故中的氧化物核燃料有两种可能的扩散方式：裂变产物能被转化为气体或以微小颗粒的形式分散分散。

Fuel behavior and post irradiation examination

Post Irradiation Examination (PIE) is the study of used nuclear materials such as nuclear fuel. It has several purposes. It is known that by examination of used fuel that the failure modes which occur during normal use (and the manner in which the fuel will behave during an accident) can be studied. In addition information is gained which enables the users of fuel to assure themselves of its quality and it also assists in the development of new fuels. After major accidents the core (or what is left of it) is normally subject to PIE to find out what happened. One site where PIE is done is the ITU which is the EU centre for the study of highly radioactive materials.

Materials in a high radiation environment (such as a reactor) can undergo unique behaviors such as swelling [2] and non-thermal creep. If there are nuclear reactions within the material (such as what happens in the fuel), the stoichiometry will also change slowly over time. These behaviors can lead to new material properties, cracking, and fission gas release.

The thermal conductivity of uranium dioxide is low; it is affected by porosity and burn-up. The burn-up results in fission products being dissolved in the lattice (such as lanthanides), the precipitation of fission products such as palladium, the formation of fission gas bubbles due to fission products such as xenon and krypton and radiation damage of the lattice. The low thermal conductivity can lead to overheating of the center part of the pellets during use. The porosity results in a decrease in both the thermal conductivity of the fuel and the swelling which occurs during use.

According to the International Nuclear Safety Center [3] the thermal conductivity of uranium dioxide can be predicted under different conditions by a series of equations.

The bulk density of the fuel can be related to the thermal conductivity

Where ρ is the bulk density of the fuel and ρ_td is the theoretical density of the uranium dioxide.

Then the thermal conductivity of the porous phase (K_f) is related to the conductivity of the perfect phase (K_o, no porosity) by the following equation. Note that s is a term for the shape factor of the holes.

K_f = K_o(1 ? p/1 + (s ? 1)p)

Rather than measuring the thermal conductivity using the traditional methods in physics such as Lees' disk, the Forbes' method or Searle's bar it is common to use a laser flash method where a small disc of fuel is placed in a furnace. After being heated to the required temperature one side of the disc is illuminated with a laser pulse, the time required for the heat wave to flow through the disc, the density of the disc, and the thickness of the disk can then be used to calculated to give the thermal conductivity.

λ = ρC_pα

• λ thermal conductivity
• ρ density
• C_p heat capacity
• α thermal diffusivity

If t_1/2 is defined as the time required for the non illuminated surface to experience half its final temperature rise then.

α = 0.1388 L²/t_1/2

• L is the thickness of the disc

For details see [4]

Radioisotope decay fuels

Radioisotope battery

The terms atomic battery, nuclear battery and radioisotope battery are used interchangely to describe a device which uses the radioactive decay to generate electricity. These systems use radioisotopes that produce low energy beta particles or sometimes alpha particles of varying energies. Low energy beta particles are needed to prevent the production of high energy penetrating Bremsstrahlung radiation that would require heavy shielding. Radioisotopes such as tritium, nickel-63, promethium-147, and technetium-99 have been tested. Plutonium-238, curium-242, curium-244 and strontium-90 have been used.

There are two main categories of atomic batteries: thermal and non-thermal. The non-thermal atomic batteries, which have many different designs, exploit charged alpha and beta particles. These designs include the direct charging generators, betavoltaics, the optoelectric nuclear battery, and the radioisotope piezoelectric generator. The thermal atomic batteries on the other hand, convert the heat from the radioactive decay to electricity. These designs include thermionic converter, thermophotovoltaic cells, alkali-metal thermal to electric converter, and the most common design, the radioisotope thermoelectric generator.

Radioisotope thermoelectric generators

A radioisotope thermoelectric generator (RTG) is a simple electrical generator which converts heat into electricity from a radioisotope using an array of thermocouples.

²³⁸Pu has become the most widely used fuel for RTGs. In the form of plutonium dioxide it has a half-life of 87.7 years, reasonable energy density and exceptionally low gamma and neutron radiation levels. Some Russian terrestrial RTGs have used ⁹⁰Sr; this isotope has a shorter half-life and a much lower energy density, but is cheaper. Early RTGs, first built in 1958 by the U.S. Atomic Energy Commission, have used ²¹⁰Po. This fuel provides phenomenally huge energy density, (a single gram of polonium-210 generates 140 watts thermal) but has limited use because of its very short half-life and gamma production and has been phased out of use in this application.

Radioisotope heater units (RHU)

Photo of a disassembled RHU

Radioisotope heater units normally provide about 1 watt of heat each, derived from the decay of a few grams of plutonium-238. This heat is given off continuously for several decades.

Their function is to provide highly localised heating of sensitive equipment (such as electronics in outer space). The Cassini–Huygens orbiter to Saturn contains 82 of these units (in addition to its 3 main RTG's for power generation). The Huygens probe to Titan contains 35 devices.

聚变核燃料

聚变核燃料包括氘（²H）、氚（³H）及氦-3（³He）等。尽管还有众多核素之间也能发生核聚变，但因为原子核所带电荷越多则需要更到的温度引发核聚变，所以仅有质量最轻的几种核素才被视为聚变核燃料。虽然核聚变的能量密度甚至比核裂变的还高，且人们已经制造出可以维持数分钟的核聚变反应堆，但将聚变核燃料用作为能源仍只在理论上可行。

第一代聚变核燃料

氘与氚都可被视作与第一代聚变核燃料。因为氘与氚所带电荷较少，所以在所有核素中它们是最易发生核聚变的。下面列举的是最常被引用的发生在第一代聚变核燃料之间的三种核反应：

²H + ³H $\rightarrow$ n (14.07 MeV) + ⁴He (3.52 MeV)

²H + ²H $\rightarrow$ n (2.45 MeV) + ³He (0.82 MeV)

²H + ²H $\rightarrow$ p (3.02 MeV) + ³H (1.01 MeV)

第二代聚变核燃料

与第一代聚变核燃料相比，第二代聚变核燃料需要更高的约束温度（confinement temperature）或更长的约束时间（confinement time），但在反应中产生的中子量较少。因为中子不带电，不受磁场约束，会被核聚变反应堆内壁吸收，使得内壁材料带上放射性，所以被视为可控核聚变中是有害副产物。第二代聚变核燃料包括氘与氦-3，虽然产物都是带电粒子，但是此代聚变核燃料也可发生不能忽略的、产生中子的副反应。

²H + ³He $\rightarrow$ p (14.68 MeV) + ⁴He (3.67 MeV)

第三代聚变核燃料

第三代聚变核燃料之间发生的反应中只产生带电粒子，且副反应可忽略。因为中子产量很低，所以使用第三代聚变核燃料的核反应堆的内壁放射性不会用明显增强。使用第三代聚变核燃料作为聚变反应堆的核燃料是可控核聚变的最终目标。在所有的第三代聚变核燃料中，氦-3具有最高的麦克斯韦反应性（Maxwellian reactivity），但是地球上氦-3的储藏量极低。

³He + ³He $\rightarrow$ 2p + ⁴He (12.86 MeV)

另一个可作为候选的无中子反应是氕-硼反应：

p + ¹¹B → 3⁴He

在合理的假设中，此反应的副反应会导致约0.1%的聚变能被中子带走。在123keV时，此反应的最佳温度约为纯氢反应的10倍，对能量约束的要求要比氘-氚反应严格500倍，但能量密度仅为氘-氚反应的0.4‰。

See also

• Global Nuclear Energy Partnership
• Lists of nuclear disasters and radioactive incidents
• Nuclear fuel bank
• Nuclear fuel cycle
• Reprocessed uranium
• Uranium market
• Integrated Nuclear Fuel Cycle Information System

外部链接

压水堆核燃料

• nucleartourist.com提供的压水堆核燃料组装示意
• 三菱燃料株式会社的压水堆核燃料棒结构示意

沸水堆核燃料

• 沸水堆核燃料棒结构示意
• nucleartourist.com提供的沸水堆相关图片

坎杜堆核燃料

• nucleartourist.com提供的坎杜堆设计基础知识
• nucleartourist.com提供的坎杜堆核燃料及反应堆相关细节
• 坎杜堆核燃料循环的发展以及坎杜堆对世界和平的潜在贡献
• 坎杜堆核燃料管理课程
• 坎杜堆核燃料棒及燃料棒束

TRISO核燃料

• 利用X射线荧光层析技术对TRISO核燃料碳化硅外壳进行的非破坏性检验
• 气透平-模块氦冷反应堆（GT-MHR）中TRISO核燃料的结构示意
• 洛斯阿拉莫斯国家实验室（LANL）提供的TRISO核燃料生产过程不同阶段介绍

QUADRISO核燃料

• QUADRISO核燃料概念设计

金属陶瓷核燃料

• 轻水堆中钼基金属陶瓷（CerMet）核燃料在锕系元素处理方面的应用

板状核燃料

• 爱达荷国家实验室（INL）的反应堆列表及高级试验反应堆（ATR）核心图片

铀氢锆堆核燃料

• 通用原子公司提供的铀氢锆堆核燃料介绍

聚变核燃料

• 高级聚变核燃料的表现