Editing Neutron radiation (section)

== Effects on materials ==
High-energy neutrons damage and degrade materials over time; bombardment of materials with neutrons creates [[collision cascade]]s that can produce [[point defect]]s and
[[dislocation]]s in the material, the creation of which is the primary driver behind microstructural changes occurring over time in materials exposed to radiation. At high neutron [[fluence]]s this can lead to [[embrittlement]] of metals and other materials, and to [[neutron-induced swelling]] in some of them. This poses a problem for nuclear reactor vessels and significantly limits their lifetime (which can be somewhat prolonged by controlled [[Annealing (metallurgy)|annealing]] of the vessel, reducing the number of the built-up dislocations). Graphite [[neutron moderator]] blocks are especially susceptible to this effect, known as [[Wigner effect]], and must be annealed periodically. The [[Windscale fire]] was caused by a mishap during such an annealing operation.

Radiation damage to materials occurs as a result of the interaction of an energetic incident particle (a neutron, or otherwise) with a lattice atom in the material. The collision causes a massive transfer of kinetic energy to the lattice atom, which is displaced from its lattice site, becoming what is known as the [[primary knock-on atom]] (PKA). Because the PKA is surrounded by other lattice atoms, its displacement and passage through the lattice results in many subsequent collisions and the creations of additional knock-on atoms, producing what is known as the collision cascade or displacement cascade. The knock-on atoms lose energy with each collision, and terminate as [[Interstitial defect|interstitials]], effectively creating a series of [[Frenkel defect]]s in the lattice. Heat is also created as a result of the collisions (from electronic energy loss), as are possibly [[nuclear transmutation|transmuted atoms]]. The magnitude of the damage is such that a single 1 [[MeV]] neutron creating a PKA in an iron lattice produces approximately 1,100 Frenkel pairs.<ref name="Lecture">Dunand, David. "Materials in Nuclear Power Generation." Materials Science & Engineering 381: Materials for Energy Efficient Technology. Northwestern University, Evanston. 3 Feb. 2015. Lecture</ref> The entire cascade event occurs over a timescale of 1 × 10<sup>−13</sup> seconds, and therefore, can only be "observed" in computer simulations of the event.<ref name="Thermal Spike Lifetime">A. Struchbery, E. Bezakova "Thermal-Spike Lifetime from Picosecond-Duration Preequilibrium Effects in Hyperfine Magnetic Fields Following Ion Implantation". 3 May. 1999.</ref>

The knock-on atoms terminate in non-equilibrium interstitial lattice positions, many of which annihilate themselves by diffusing back into neighboring vacant lattice sites and restore the ordered lattice. Those that do not or cannot leave vacancies, which causes a local rise in the vacancy concentration far above that of the equilibrium concentration. These vacancies tend to migrate as a result of [[Thermal transpiration|thermal diffusion]] towards vacancy sinks (i.e., [[grain boundaries]], [[dislocations]]) but exist for significant amounts of time, during which additional high-energy particles bombard the lattice, creating collision cascades and additional vacancies, which migrate towards sinks. The main effect of irradiation in a lattice is the significant and persistent flux of defects to sinks in what is known as the [[defect wind]]. Vacancies can also annihilate by combining with one another to form [[pinning point|dislocation loops]] and later, [[crystallographic defect|lattice voids]].<ref name="Lecture" />

The collision cascade creates many more vacancies and interstitials in the material than equilibrium for a given temperature, and [[diffusivity]] in the material is dramatically increased as a result. This leads to an effect called [[radiation-enhanced diffusion]], which leads to microstructural evolution of the material over time. The mechanisms leading to the evolution of the microstructure are many, may vary with temperature, flux, and fluence, and are a subject of extensive study.<ref name ="Radiation Effects in Nuclear Ceramics">{{cite journal|title=Radiation Effects in Nuclear Ceramics|first1=L.|last1=Thomé|first2=S.|last2=Moll|first3=A.|last3=Debelle|first4=F.|last4=Garrido|first5=G.|last5=Sattonnay|first6=J.|last6=Jagielski|date=1 June 2018|journal=Advances in Materials Science and Engineering|volume=2012|pages=1–13|doi=10.1155/2012/905474|doi-access=free}}</ref>

* [[Radiation-induced segregation]] results from the aforementioned flux of vacancies to sinks, implying a flux of lattice atoms away from sinks; but not necessarily in the same proportion to alloy composition in the case of an alloyed material. These fluxes may therefore lead to depletion of alloying elements in the vicinity of sinks. For the flux of interstitials introduced by the cascade, the effect is reversed: the interstitials diffuse toward sinks resulting in alloy enrichment near the sink.<ref name="Lecture" />
* [[pinning point|Dislocation loops]] are formed if vacancies form clusters on a lattice plane. If these vacancy concentration expand in three dimensions, a [[Vacuum|void]] forms. By definition, voids are under vacuum, but may became gas-filled in the case of [[alpha particle|alpha-particle radiation]] (helium) or if the gas is produced as a result of [[nuclear transmutation|transmutation reactions]]. The void is then called a bubble, and leads to dimensional instability (neutron-induced swelling) of parts subject to radiation. Swelling presents a major long-term design problem, especially in reactor components made out of stainless steel.<ref name="Voids in Irradiated Stainless Steel">{{cite journal|title=Voids in Irradiated Stainless Steel|first1=C.|last1=CAWTHORNE|first2=E. J.|last2=FULTON|date=1 November 1967|journal=Nature|volume=216|issue=5115|pages=575–576|doi=10.1038/216575a0|bibcode=1967Natur.216..575C|s2cid=4238714}}</ref> Alloys with crystallographic [[isotropy]], such as [[Zircaloy]]s are subject to the creation of dislocation loops, but do not exhibit void formation. Instead, the loops form on particular lattice planes, and can lead to [[irradiation-induced growth]], a phenomenon distinct from swelling, but that can also produce significant dimensional changes in an alloy.<ref name="Effects of Neutron Radiation on Microstructure and Properties of Zircaloy">Adamson, R. "Effects of Neutron Radiation on Microstructure and the Properties of Zircaloy" 1977. 08 Feb. 2015.</ref>
*Irradiation of materials can also induce [[phase transformation]]s in the material: in the case of a [[solid solution]], the solute enrichment or depletion at sinks radiation-induced segregation can lead to the precipitation of new phases in the material.<ref name="Neutron irradiation performance of Zircaloy-4 under research reactor operating conditions">Hyun Ju Jin, Tae Kyu Kim. "Neutron irradiation performance of Zircaloy-4 under research reactor operating conditions." Annals of Nuclear Energy. 13 Sept. 2014 Web. 08 Feb. 2015.</ref>

The mechanical effects of these mechanisms include [[irradiation hardening]], [[embrittlement]], [[creep (deformation)|creep]], and [[stress corrosion cracking|environmentally-assisted cracking]]. The defect clusters, dislocation loops, voids, bubbles, and precipitates produced as a result of radiation in a material all contribute to the strengthening and [[embrittlement]] (loss of [[ductility]]) in the material.<ref name="Effect of Irradiation">{{cite book|chapter-url=http://www.astm.org/DIGITAL_LIBRARY/STP/PAGES/STP33683S.htm|title=Effects of Radiation on Structural Materials|first=CJ|last=Baroch|publisher=ASTM International|website=astm.org|pages=129–129–14|doi=10.1520/STP33683S|chapter=Effect of Irradiation at 130, 650, and 775°F on Tensile Properties of Zircaloy-4 at 70, 650, and 775°F|year=1975|doi-broken-date=25 March 2025 |isbn=978-0-8031-0539-3}}</ref> Embrittlement is of particular concern for the material comprising the reactor pressure vessel, where as a result the energy required to fracture the vessel decreases significantly. It is possible to restore ductility by annealing the defects out, and much of the life-extension of nuclear reactors depends on the ability to safely do so. [[Creep (deformation)|Creep]] is also greatly accelerated in irradiated materials, though not as a result of the enhanced diffusivities, but rather as a result of the interaction between lattice stress and the developing microstructure. Environmentally-assisted cracking or, more specifically, [[irradiation assisted stress corrosion cracking|irradiation-assisted stress corrosion cracking]] (IASCC) is observed especially in alloys subject to neutron radiation and in contact with water,  caused by [[hydrogen embrittlement|hydrogen absorption]] at crack tips resulting from [[radiolysis]] of the water, leading to a reduction in the required energy to propagate the crack.<ref name="Lecture" />