Source Neutrons and External Source of Neutrons

The term “source neutrons” refers to neutrons other than prompt or delayed fission neutrons, and in general, they come from sources other than neutron-induced fission. These neutrons are very important during reactor startup and shutdown operations when the reactor is subcritical because they allow monitoring the subcriticality of a reactor, usually via source range excore neutron detectors.

Do not confuse terms source neutrons and external source of neutrons. The source neutrons can be classified by the place of origin:

• Internal Source of Neutrons. A very important source of neutrons arises from the nature of nuclear fuel. Nuclear fuel, especially when irradiated, produces source neutrons that significantly monitor the subcriticality of a reactor. Irradiated nuclear fuel contains almost an entire periodic table of elements, and there is a very high radiation background inside fuel elements. Therefore, there are very many channels, how these neutrons are produced. The amount of fission products present in the fuel elements depends on how long the reactor has been operated before shutdown and at which power level has been the reactor operated before shutdown. These neutrons can be classified by type of reaction as:
  • Spontaneous fission of certain nuclides. Some nuclides undergo a process of spontaneous fission. Spontaneous fission is, in fact, a form of radioactive decay that is found only in very heavy chemical elements (especially transuranic elements). For example, ²⁴⁰Pu has a very high rate of spontaneous fission. For high burnup fuel, the neutrons are provided  predominantly by the spontaneous fission of curium nuclei (Cm-242 and Cm-244)
  • (α,n) reactions. In certain light isotopes, the ‘last’ neutron in the nucleus is weakly bound and is released when the compound nucleus formed following α-particle bombardment decays. For example, the bombardment of beryllium by α-particles leads to the production of neutrons by the following exothermic reaction: ⁴He + ⁹Be→¹²C + n + 5.7 MeV. This reaction yields a weak source of neutrons with an energy spectrum resembling that from a fission source. Another very important source is (α,n) reaction with oxygen-18. This element is present in the uranium oxide fuel. The neutron-producing reaction is:  18O(α,n)21Ne. Alpha particles are commonly emitted by all of the heavy radioactive nuclei occurring in fuel (uranium), as well as the transuranic elements (neptunium, plutonium, or americium).
  • (γ,n) reactions – photoneutrons. Neutrons can also be produced in (γ, n) reactions, and therefore they are usually referred to as photoneutrons. A high-energy photon (gamma-ray) can, under certain conditions, eject a neutron from a nucleus. It occurs when its energy exceeds the binding energy of the neutron in the nucleus. The lowest threshold has ⁹Be with 1.666 MeV and ²D with 2.226 MeV. Therefore photoneutrons are very important, especially in nuclear reactors with D₂O moderator (CANDU reactors) or Be reflectors (some experimental reactors).
• External Source of Neutrons. Sometimes source neutrons must be artificially added to the system. An external source of neutrons contains a material that emits neutrons and cladding to provide a barrier between the reactor coolant and the material. External sources are usually loaded directly into the reactor core (e.g., guide thimble tubes).
  • Primary Source of Neutrons. The primary source of neutrons does not need to be irradiated to produce neutrons. These sources can be used especially in the case of a first core (i.e., a core consisting of fresh fuel only). The primary source of neutrons is based on the spontaneous fission reaction. The most commonly used spontaneous fission source is the radioactive isotope californium-252. Cf-252 and other spontaneous fission neutron sources are produced by irradiating uranium or another transuranic element in a nuclear reactor, where neutrons are absorbed in the starting material and its subsequent reaction products, transmuting the starting material into the SF isotope. (α,n) reactions can also be used to produce neutrons. The bombardment of beryllium by α-particles leads to the production of neutrons by the following exothermic reaction: ⁴He + ⁹Be→¹²C + n + 5.7 MeV. This reaction yields a weak source of neutrons with an energy spectrum resembling that from a fission source and is used nowadays in portable neutron sources. Radium, plutonium, or americium can be used as an α-emitter.
  • Secondary Source of Neutrons. The secondary source of neutrons must be irradiated to produce neutrons. These sources usually contain two components. The second component is a neutron emitter element, for example, ⁹Be, which has the lowest threshold (1.666 MeV) for neutron emission. The first component is a material, which can be irradiated by fission neutrons (e.g., antimony) and then emits particle capable of knocking out a neutron from the second component. The (γ,n) source that uses antimony-124 as the gamma emitter is characterized in the following endothermic reaction. The antimony-beryllium source produces nearly monoenergetic neutrons with the dominant peak at 24keV.

¹²⁴Sb→¹²⁴Te + β− + γ

γ + ⁹Be→⁸Be + n – 1.66 MeV

In mechanical construction, the primary and secondary sources are similar to a control rod. Both types of source rods are clad in stainless steel.

Sb-Be Source – Antimony-Beryllium Source

Sb-Be neutron source is a typical external neutron source used in commercial nuclear reactors. It is usually loaded into fuel assemblies near the periphery of the core because the source-range excore detectors are installed outside the pressure vessel. Sb-Be source is a two-component source and, prior to irradiation, contains natural antimony (57.4% of ¹²¹Sb and 42.6% of ¹²³Sb) and natural beryllium (100% of ⁹Be). The first component is a source of strong gamma rays, while the second is a neutron emitter element.

Thus the Sb-Be source is based on (γ,n) reaction (i.e., it emits photoneutrons). The neutron flux inside the core activates (σ_a = 0.02 barn) the isotope ¹²³Sb resulting in ¹²⁴Sb. ¹²⁴Sb decays (with a half-life of 60.2 days) via beta decay to ¹²⁴Te. The decay scheme of ¹²⁴Sb shows two relevant groups of γ-ray energies, namely 1691 keV and 2091 keV, with absolute intensities of 0.484 and 0.057 per decay, respectively.

These γ-rays have sufficient energy to knock out a neutron from the second component, ⁹Be, which has the lowest threshold (1.66 MeV) for neutron emission. The (γ,n) source that uses antimony-124 as the gamma emitter is characterized in the following endothermic reaction.

¹²⁴Sb→¹²⁴Te + β− + γ

γ + ⁹Be→⁸Be + n – 1.66 MeV

The antimony-beryllium source produces nearly monoenergetic neutrons with the dominant peak at 23keV. Using the laws of energy and momentum conservation, one can derive that the 1691 keV and 2091 keV gamma rays produce two groups of neutrons:

• 23 keV (~97%)
• 378 keV  (~3%)

Sb-Be sources have three main disadvantages:

• They have an extremely high photon to neutron ratio, complicating work with such sources.
• The yield of neutrons is significantly lower than for (α,n) sources. On the other hand, (α,n) sources contain transuranic elements such as americium which can be converted into fissile isotope ^242mAm. These sources are not appropriate for commercial reactors.
• When loaded into a reactor, the Sb-Be source may contribute to the production of tritium. The most important source (due to releases of tritiated water) of tritium in nuclear power plants stems from boric acid. Still, it can also be produced from beryllium via the following sequence of reactions:
  • ⁹Be(n,α)⁶He
  • ⁶He →⁶Li + e^–   (β^–;   T_1/2=0.8s)
  • ⁶Li(n,α)³H

The geometry of Source Neutrons Assemblies

Since at PWRs, the source range neutron detectors are usually placed outside the reactor (excore). A source neutrons assemblies should be placed at least a few migration lengths from the core-periphery. The main reason is that source range detectors should not register the source neutrons primarily.

Source Neutrons and Subcritical Multiplication

Source neutrons play an important role in reactor safety, especially during shutdown state and reactor startup. Without source neutrons, there would be no subcritical multiplication, and the neutron population in the subcritical system would gradually approach zero. That means each neutron generation would have fewer neutrons than the previous one because k_eff is less than 1.0.

With source neutrons, the population remains at levels that can be measured by the source range of excore neutron detectors so that operators can always monitor how fast the neutron population is changing (can always monitor the reactivity of subcritical reactor). Note that if neutrons and fissionable material are present in the subcritical reactor, fission will occur (even a deep subcritical reactor will always produce a small number of fissions).

The source neutrons enter the life cycle and experience the same environment that fission neutrons experience. It must be noted source neutrons are produced at different energies (e.g., 24keV for Sb-Be source), usually below energies of fission neutrons.

When the reactor is made subcritical after operating at a critical state, the neutron population at first undergoes a prompt drop due to a rapid decrease in prompt neutrons. After a short time begins to decrease exponentially with a period corresponding to decay of the longest-lived delayed neutron precursors (i.e., ~ the 80s). With source neutrons, the neutron flux stabilizes itself at a corresponding level, determined by source strength, S, and the multiplication factor, k_eff.

For a defined source strength of n₀ neutrons per neutron generation, the neutron population (i.e., the neutron flux) is given by:

where n₀ is source neutrons (0. generation) and n₀k_effⁱ are neutrons from i-th generation. As i goes to infinity, the sum of this geometric series is (for k_eff < 1):

As can be seen, the neutron population of a subcritical reactor with source neutrons does not drop to zero, and the neutron population stabilizes itself at level n, which is equal to source multiplied by factor M.