What are Neutrons

What is Neutron

A neutron is one of the subatomic particles that make up matter. In the universe, neutrons are abundant, making up more than half of all visible matter. It has no electric charge and a rest mass equal to 1.67493 × 10−27 kg—marginally greater than that of the proton but nearly 1839 times greater than that of the electron. The neutron has a mean square radius of about 0.8×10−15 m, or 0.8 fm, and it is a spin-½ fermion.

The neutrons exist in the nuclei of typical atoms, along with their positively charged counterparts, the protons. Neutrons and protons, commonly called nucleons, are bound together in the atomic nucleus, where they account for 99.9 percent of the atom’s mass. Research in high-energy particle physics in the 20th century revealed that neither the neutron nor the proton is not the smallest building block of matter. Protons and neutrons have also their structure. Inside the protons and neutrons, we find true elementary particles called quarks. Within the nucleus, protons and neutrons are bound together through the strong force, a fundamental interaction that governs the behaviour of the quarks that make up the individual protons and neutrons.

A nuclear stability is determined by the competition between two fundamental interactions. Protons and neutrons are attracted each other via strong force. On the other hand protons repel each other via the electric force due to their positive charge. Therefore neutrons within the nucleus act somewhat like nuclear glue, neutrons attract each other and protons , which helps offset the electrical repulsion between protons. There are only certain combinations of neutrons and protons, which forms stable nuclei. For example, the most common nuclide of the common chemical element lead (Pb) has 82 protons and 126 neutrons.

Nuclear binding energy curve.
Nuclear binding energy curve.
Source: hyperphysics.phy-astr.gsu.edu

Because of the strength of the nuclear force at short distances, the nuclear binding energy (the energy required to disassemble a nucleus of an atom into its component parts) of nucleons is more than seven orders of magnitude larger than the electromagnetic energy binding electrons in atoms. Nuclear reactions (such as nuclear fission or nuclear fusion) therefore have an energy density that is more than 10 000 000x that of chemical reactions.
Knowledge of the behaviour and properties of neutrons is essential to the production of nuclear power. Shortly after the neutron was discovered in 1932, it was quickly realized that neutrons might act to form a nuclear chain reaction. When nuclear fission was discovered in 1938, it became clear that, if a fission reaction produced free neutrons, each of these neutrons might cause further fission reaction in a cascade known as a chain reaction. Knowledge of cross-sections (the key parameter representing probability of interaction between a neutron and a nucleus) became crutial for design of reactor cores and the first nuclear weapon (Trinity, 1945).

Discovery of the Neutron
The story of the discovery of the neutron and its properties is central to the extraordinary developments in atomic physics that occurred in the first half of the 20th century. The neutron was discovered in 1932 by the English physicist James Chadwick, but since the time of Ernest Rutherford it had been known that the atomic mass number A of nuclei is a bit more than twice the atomic number Z for most atoms and that essentially all the mass of the atom is concentrated in the relatively tiny nucleus. The Rutherford’s model for the atom in 1911 claims that atoms have their mass and positive charge concentrated in a very small nucleus.
Discovery of the Neutron
The alpha particles emitted from polonium fell on certain light elements, specifically beryllium, an unusually penetrating radiation is produced.
Source: dev.physicslab.org
Chadwicks chamber.
Chadwick’s neutron chamber containing parallel disks of radioactive polonium and beryllium. Radiation is emitted from an aluminium window at the chamber’s end.
Source: imgkid.com

An experimental breakthrough came in 1930 with the observation by Bothe and Becker. They found that if the very energetic alpha particles emitted from polonium fell on certain light elements, specifically beryllium, boron, or lithium, an unusually penetrating radiation was produced. Since this radiation was not influenced by an electric field (neutrons have no charge), they presumed it was gamma rays (but much more penetrating). It was shown (Curie and Joliot) that when a paraffin target with this radiation is bombarded, it ejected protons with energy about 5.3 MeV. Paraffin is high in hydrogen content, hence offers a target dense with protons (since neutrons and protons have almost equal mass, protons scatter energetically from neutrons).These experimental results were difficult to interpret. James Chadwick was able to prove that the neutral particle could not be a photon by bombarding targets other than hydrogen, including nitrogen, oxygen, helium and argon. Not only were these inconsistent with photon emission on energy grounds, the cross-section for the interactions was orders of magnitude greater than that for Compton scattering by photons. In Rome, the young physicist Ettore Majorana suggested that the manner in which the new radiation interacted with protons required a new neutral particle.

The task was that of determining the mass of this neutral particle. James Chadwick chose to bombard boron with alpha particles and analyze the interaction of the neutral particles with nitrogen. These particlular targets were chosen partly because the masses of boron and nitrogen were well known. Using kinematics, Chadwick was able to determine the velocity of the protons. Then through conservation of momentum techniques, he was able to determine that the mass of the neutral radiation was almost exactly the same as that of a proton. In 1932, Chadwick proposed that the neutral particle was Rutherford’s neutron. In 1935, he was awarded the Nobel Prize for his discovery.

See also: Discovery of the Neutron

Structure of the Neutron

Quark structure of the Neutron
The quark structure of the neutron. The color assignment of individual quarks is arbitrary, but all three colors must be present. Forces between quarks are mediated by gluons.

Neutrons and protons are classified as hadrons, subatomic particles that are subject to the strong force and as baryons since they are composed of three quarks. The neutron is a composite particle made of two down quarks with charge −⅓  e and one up quark with charge +⅔ e. Since the neutron has no net electric charge, it is not affected by eletric forces, but the neutron does have a slight distribution of electric charge within it. This results in non-zero magnetic moment (dipole moment) of the neutron. Therefore the neutron interacts also via electromagnetic interaction, but much weaker than the proton.

The mass of the neutron is 939.565 MeV/c2, whereas the mass of the three quarks is only about 12 MeV/c2 (only about 1% of the mass-energy of the neutron). Like the proton, most of mass (energy) of the neutron is in the form of the strong nuclear force energy (gluons). The quarks of the neutron are held together by gluons, the exchange particles for the strong nuclear force. Gluons carry the color charge of the strong nuclear force.

See also: Structure of the Neutron

Properties of the Neutron

Key properties of neutrons are summarized below:

  • Mean square radius of a neutron is ~ 0.8 x 10-15m (0.8 fermi)
  • The mass of the neutron is 939.565 MeV/c2
  • Neutrons are ½ spin particles – fermionic statistics
  • Neutrons are neutral particles – no net electric charge.
  • Neutrons have non-zero magnetic moment.
  • Free neutrons (outside a nucleus) are unstable and decay via beta decay. The decay of the neutron involves the weak interaction and is associated with a quark transformation (a down quark is converted to an up quark).
  • Mean lifetime of a free neutron is 882 seconds (i.e. half-life is 611 seconds ).
  • A natural neutron background of free neutrons exists everywhere on Earth and it is caused by muons produced in the atmosphere, where high energy cosmic rays collide with particles of Earth’s atmosphere.
  • Neutrons cannot directly cause ionization. Neutrons ionize matter only indirectly.
  • Neutrons can travel hundreds of feet in air without any interaction. Neutron radiation is highly penetrating.
  • Neutrons trigger the nuclear fission.
  • The fission process produces free neutrons (2 or 3).
  • Thermal or cold neutrons have the wavelengths similar to atomic spacings. They can be used in neutron diffraction experiments to determine the atomic and/or magnetic structure of a material.

See also: Properties of the Neutron

Neutron Energy
Free neutrons can be classified according to their kinetic energy. This energy is usually given in electron volts (eV). The term temperature can also describe this energy representing thermal equilibrium between a neutron and a medium with a certain temperature.

Classification of free neutrons according kinetic energies

  • Cold Neutrons (0 eV; 0.025 eV). Neutrons in thermal equilibrium with very cold surroundings such as liquid deuterium. This spectrum is used for neutron scattering experiments.
  • Thermal Neutrons. Neutrons in thermal equilibrium with a surrounding medium. Most probable energy at 20°C (68°F) for Maxwellian distribution is 0.025 eV (~2 km/s). This part of neutron’s energy spectrum constitutes most important part of spectrum in thermal reactors.
  • Epithermal Neutrons (0.025 eV; 0.4 eV). Neutrons of kinetic energy greater than thermal. Some of reactor designs operates with epithermal neutron’s spectrum. This design allows to reach higher fuel breeding ratio than in thermal reactors.
  • Cadmium cut-off energy
    Neutrons of kinetic energy below the cadmium cut-off energy (~0.5 eV) are strongly absorbed by 113-Cd.
    Source: JANIS (Java-based nuclear information software) www.oecd-nea.org/janis/

    Cadmium Neutrons (0.4 eV; 0.5 eV). Neutrons of kinetic energy below the cadmium cut-off energy. One cadmium isotope, 113Cd, absorbs neutrons strongly only if they are below ~0.5 eV (cadmium cut-off energy).

  • Epicadmium Neutrons (0.5 eV; 1 eV). Neutrons of kinetic energy above the cadmium cut-off energy. These neutrons are not absorbed by cadmium.
  • Slow Neutrons (1 eV; 10 eV).
  • Resonance Neutrons (10 eV; 300 eV). The resonance neutrons are called resonance for their special bahavior. At resonance energies the cross-sections can reach peaks more than 100x higher as the base value of cross-section. At this energies the neutron capture significantly exceeds a probability of fission. Therefore it is very important (for thermal reactors) to quickly overcome this range of energy and operate the reactor with thermal neutrons resulting in increase of probability of fission.
  • Intermediate Neutrons (300 eV; 1 MeV).
  • Fast Neutrons (1 MeV; 20 MeV). Neutrons of kinetic energy greater than 1 MeV (~15 000 km/s) are usually named fission neutrons. These neutrons are produced by nuclear processes such as nuclear fission or (ɑ,n) reactions. The fission neutrons have a Maxwell-Boltzmann distribution of energy with a mean energy (for 235U fission) 2 MeV. Inside a nuclear reactor the fast neutrons are slowed down to the thermal energies via a process called neutron moderation.
  • Relativistic Neutrons (20 MeV; ->)
Neutron energies in thermal reactor
Distribution of kinetic energies of neutrons in the thermal reactor. The fission neutrons (fast flux) are immediately slowed down to the thermal energies via a process called neutron moderation.
Source: serc.carleton.edu

The reactor physics does not need this fine division of neutron energies. The neutrons can be roughly (for purposes of reactor physics) divided into three energy ranges:

  • Thermal neutrons (0.025 eV – 1 eV).
  • Resonance neutrons (1 eV – 1 keV).
  • Fast neutrons (1 keV – 10 MeV).

Even most of reactor computing codes use only two neutron energy groups:

  • Slow neutrons group (0.025 eV – 1 keV).
  • Fast neutrons group (1 keV – 10 MeV).

See also: Neutron Energy

Interactions of Neutrons with Matter

Neutron - Nuclear ReactionsNeutrons are neutral particles, therefore they travel in straight lines, deviating from their path only when they actually collide with a nucleus to be scattered into a new direction or absorbed. Neither the electrons surrounding (atomic electron cloud) a nucleus nor the electric field caused by a positively charged nucleus affect a neutron’s flight. In short, neutrons collide with nuclei, not with atoms. A very descriptive feature of the transmission of neutrons through bulk matter is the mean free path length (λ – lambda), which is the mean distance a neutron travels between interactions. It can be calculated from following equation:

λ=1/Σ

Neutrons may interact with nuclei in one of following ways:

Neutron Cross-section
Neutron cross-section
Typical cross-sections of fission material. Slowing down neutrons results in increase of probability of interaction (e.g. fission reaction).

The extent to which neutrons interact with nuclei is described in terms of quantities known as cross-sections. Cross-sections are used to express the likelihood of particular interaction between an incident neutron and a target nucleus. It must be noted this likelihood do not depend on real target dimensions. In conjunction with the neutron flux, it enables the calculation of the reaction rate, for example to derive the thermal power of a nuclear power plant. The standard unit for measuring the microscopic cross-section (σ-sigma) is the barn, which is equal to 10-28 m2. This unit is very small, therefore barns (abbreviated as “b”) are commonly used. The microscopic cross-section can be interpreted as the effective ‘target area’ that a nucleus interacts with an incident neutron.

A macroscopic cross-section is derived from microscopic and the material density:

 Σ=σ.N

 Here σ, which has units of m2, is referred to as the microscopic cross-section. Since the units of N (nuclei density) are nuclei/m3, the macroscopic cross-section Σ have units of m-1, thus in fact is an incorrect name, because it is not a correct unit of cross-sections.

Neutron cross-sections constitute a key parameters of nuclear fuel. Neutron cross-sections  must be calculated for fresh fuel assemblies usually in two-Dimensional models of the fuel lattice.

 The neutron cross-section is variable and depends on:

  • Target nucleus (hydrogen, boron, uranium, etc.) Each isotop has its own set of cross-sections.
  • Type of the reaction (capture, fission, etc.). Cross-sections are different for each nuclear reaction.
  • Neutron energy (thermal neutron, resonance neutron, fast neutron). For a given target and reaction type, the cross-section is strongly dependent on the neutron energy. In the common case, the cross section is usually much larger at low energies than at high energies. This is why most nuclear reactors use a neutron moderator to reduce the energy of the neutron and thus increase the probability of fission, essential to produce energy and sustain the chain reaction.
  • Target energy (temperature of target material – Doppler broadening) This dependency is not so significant, but the target energy strongly influences inherent safety of nuclear reactors due to a Doppler broadening of resonances.

See also: JANIS (Java-based nuclear information software) 

See also: Interactions of Neutrons with Matter

See also: Neutron cross-section

Law 1/v

1/v Law
For thermal neutrons (in 1/v region), absorption cross sections increases as the velocity (kinetic energy) of the neutron decreases.
Source: JANIS 4.0

For thermal neutrons (in 1/v region),  absorption cross-sections increases as the velocity (kinetic energy) of the neutron decreases. Therefore the 1/v Law can be used to determine shift in absorbtion cross-section, if the neutron is in equilibrium with a surrounding medium. This phenomenon is due to the fact the nuclear force between the target nucleus and the neutron has a longer time to interact.

\sigma_a \sim \frac{1}{v}}} \sim \frac{1}{\sqrt{E}}}}} \sim \frac{1}{\sqrt{T}}}}}

This law is aplicable only for absorbtion cross-section and only in the 1/v region.

Example of cross- sections in 1/v region:

The absorbtion cross-section for 238U at 20°C = 293K (~0.0253 eV) is:

\sigma_a(293K) = 2.68b .

The absorbtion cross-section for 238U at 1000°C = 1273K is equal to:

Neutron Cross-section - 1-v law

This cross-section reduction is caused only due to the shift of temperature of surrounding medium.

Resonance neutron capture

Resonance peaks for radiative capture of U238.
Resonance peaks for radiative capture of U238. At resonance energies the probability of capture can be more than 100x higher as the base value.
Source: JANIS program

Absorption cross section is often highly dependent on neutron energy. Note that the nuclear fission produces neutrons with a mean energy of 2 MeV (200 TJ/kg, i.e. 20,000 km/s). The neutron can be roughly divided into three energy ranges:

  • Fast neutron. (10MeV – 1keV)
  • Resonance neutron (1keV – 1eV)
  • Thermal neutron. (1eV – 0.025eV)

The resonance neutrons are called resonance for their special bahavior. At resonance energies the cross-section can reach peaks more than 100x higher as the base value of cross-section. At this energies the neutron capture significantly exceeds a probability of fission. Therefore it is very important (for thermal reactors) to quickly overcome this range of energy and operate the reactor with thermal neutrons resulting in increase of probability of fission.

Doppler broadening

 

Doppler effect
Doppler effect improves reactor stability. Broadened resonance (heating of a fuel) results in a higher probability of absorbtion, thus causes negative reactivity insertion (reduction of reactor power).

A Doppler broadening of resonances is very important phanomenon, which improves reactor stability. The prompt temperature coefficient of most thermal reactors is negative, owing to an nuclear Doppler effect. Although the absorbtion cross-section depends significantly on incident neutron energy, the shape of the cross-section curve depends also on target temperature.

Nuclei are located in atoms which are themselves in continual motion owing to their thermal energy. As a result of these thermal motions neutrons impinging on a target appears to the nuclei in the target to have a continuous spread in energy. This, in turn, has an effect on the observed shape of resonance. The resonance becomes shorter and wider than when the nuclei are at rest.

Although the shape of a resonance changes with temperature, the total area under the resonance remains essentially constant. But this does not imply constant neutron absorbtion. Despite the constant area under resonance, a resonance integral, which determines the absorbtion, increases with increasing target temperature. This, of course, decreases coefficient k (negative reactivity is inserted).

Typical cross-sections of materials in the reactor

Following table shows neutron cross-sections of the most common isotopes of reactor core.

Table of cross-sections
Table of cross-sections

Types of neutron-nuclear reactions

Elastic Scattering Reaction
Generally, a neutron scattering reaction occurs when a target nucleus emits a single neutron after a neutron-nucleus interaction. In an elastic scattering reaction between a neutron and a target nucleus, there is no energy transferred into nuclear excitation.
Inelastic Scattering Reaction
In an inelastic scattering reaction between a neutron and a target nucleus some energy of the incident neutron is absorbed to the recoiling nucleus and the nucleus remains in the excited state. Thus while momentum is conserved in an inelastic collision, kinetic energy of the “system” is not conserved.
Neutron Absorption
The neutron absorption reaction is the most important type of reactions that take place in a nuclear reactor. The absorption reactions are reactions, where the neutron is completely absorbed and compound nucleus is formed. This is the very important feature, because the mode of decay of such compound nucleus does not depend on the way the compound nucleus was formed. Therefore a variety of emissions or decays may follow. The most important absorption reactions are divided by the exit channel into two following reactions:
  • Radiative Capture. Most absorption reactions result in the loss of a neutron coupled with the production of one or more gamma rays. This is referred to as a capture reaction, and it is denoted by σγ.
  • Neutron-induced Fission Reaction. Some nuclei (fissionable nuclei) may undergo a fission event, leading to two or more fission fragments (nuclei of intermediate atomic weight) and a few neutrons. In a fissionable material, the neutron may simply be captured, or it may cause nuclear fission. For fissionable materials we thus divide the absorption cross section as σa = σγ + σf.
Radiative Capture
The neutron capture is one of the possible absorption reactions that may occur. In fact, for non-fissionable nuclei it is the only possible absorption reaction. Capture reactions result in the loss of a neutron coupled with the production of one or more gamma rays. This capture reaction is also referred to as a radiative capture or (n, γ) reaction, and its cross-section is denoted by σγ.

The radiative capture is a reaction, in which the incident neutron is completely absorbed and compound nucleus is formed. The compound nucleus then decays to its ground state by gamma emission. This process can occur at all incident neutron energies, but the probability of the interaction strongly depends on the incident neutron energy and also on the target energy (temperature). In fact the energy in the center-of-mass system determines this probability.

Nuclear Fission
Nuclear fission is a nuclear reaction in which the nucleus of an atom splits into smaller parts (lighter nuclei). The fission process often produces free neutrons and photons (in the form of gamma rays), and releases a large amount of energy. In nuclear physics, nuclear fission is either a nuclear reaction or a radioactive decay process. The case of decay process is called spontaneous fission and it is very rare process.
Neutron Emission
Although the neutron emission is usually associated with nuclear decay, it must be also mentioned in connection with neutron nuclear reactions. Some neutrons interacts with a target nucleus via a compound nucleus. Among these compound nucleus reactions are also reactions, in which a neutron is ejected from nucleus and they may be referred to as neutron emission reactions. The point is that compound nuclei lose its excitation energy in a way, which is identical to the radioactive decay. Very important feature is the fact the mode of decay of compound nucleus does not depend on the way the compound nucleus was formed.
Charged Particle Ejection
Charged particle reactions are usually associated with formation of a compound nucleus, which is excited to a high energy level, that such compound nucleus can eject a new charged particle while the incident neutron remains in the nucleus. After the new particle is ejected, the remaining nucleus is completely changed, but may or may not exist in an excited state depending upon the mass-energy balance of the reaction. This type of reaction is more common for charged particles as incident particles (such as alpha particles, protons, and so on).

The case of neutron-induced charged particle reactions is not so common, but there are some neutron-induced charged particle reactions, that are of importance in the reactivity control and also in the detection of neutrons.

Detection of Neutrons

Since the neutrons are electrically neutral particles, they are mainly subject to strong nuclear forces but not to electric forces. Therefore neutrons are not directly ionizing and they have usually to be converted into charged particles before they can be detected. Generally every type of neutron detector must be equipped with converter (to convert neutron radiation to common detectable radiation) and one of the conventional radiation detectors (scintillation detector, gaseous detector, semiconductor detector, etc.).

Neutron converters

Two basic types of neutron interactions with matter are for this purpose available:

  • Elastic scattering. The free neutron can be scattered by a nucleus, transferring some of its kinetic energy to the nucleus. If the neutron has enough energy to scatter off nuclei the recoiling nucleus ionizes the material surrounding the converter. In fact, only hydrogen and helium nuclei are light enough for practical application. Charge produced in this way can be collected by the conventional detector to produce a detected signal. Neutrons can transfer more energy to light nuclei. This method is appropriate for detecting fast neutrons (fast neutrons do not have high cross-section for absorption) allowing detection of fast neutrons without a moderator.
  • Neutron absorption. This is a common method allowing detection of neutrons of entire energy spectrum. This method is is based on variety of absorption reactions (radiative capture, nuclear fission, rearrangement reactions, etc.). The neutron is here absorbed by target material (converter) emitting secondary particles such as protons, alpha particles, beta particles, photons (gamma rays) or fission fragments. Some reactions are threshold reactions (requiring a minimum energy of neutrons), but most of reactions occurs at epithermal and thermal energies. That means the moderation of fast neutrons is required leading in poor energy information of the neutrons. Most common nuclei for the neutron converter material are:
    • 10B(n,α). Where the neutron capture cross-section for thermal neutrons is σ = 3820 barns and the natural boron has abundance of 10B 19,8%.
    • 3He(n,p). Where the neutron capture cross-section for thermal neutrons is σ = 5350 barns and the natural helium has abundance of 3He 0.014%.
    • 6Li(n,α). Where the neutron capture cross-section for thermal neutrons is σ = 925 barns and the natural lithium has abundance of 6Li 7,4%.
    • 113Cd(n,ɣ). Where the neutron capture cross-section for thermal neutrons is σ = 20820 barns and the natural cadmium has abundance of 113Cd 12,2%.
    • 235U(n,fission). Where the fission cross-section for thermal neutrons is σ = 585 barns and the natural uranium has abundance of 235U 0.711%. Uranium as a converter produces fission fragments which are heavy charged particles. This have significant advantage. The heavy charged particles (fission fragments) create a high output signal, because the fragments deposit a large amount of energy in a detector sensitive volume. This allows an easy discrimination of the background radiation (e.i. gamma radiation). This important feature can be used for example in a nuclear reactor power measurement, where the neutron field is accompanied  by a significant gamma background.

See also: Detection of Neutrons

Free Neutron
Free Neutron
The free neutron decays into a proton, an electron, and an antineutrino with a half-life of about 611 seconds (10.3 minutes).
Source: scienceblogs.com

A free neutron is a neutron that is not bounded in a nucleus. The free neutron is, unlike a bounded neutron, subject to radioactive beta decay.

It decays into a proton, an electron, and an antineutrino (the antimatter counterpart of the neutrino, a particle with no charge and little or no mass). A free neutron will decay with a half-life of about 611 seconds (10.3 minutes). This decay involves the weak interaction and is associated with a quark transformation (a down quark is converted to an up quark). The decay of the neutron is a good example of the observations which led to the discovery of the neutrino. Because it decays in this manner, the neutron does not exist in nature in its free state, except among other highly energetic particles in cosmic rays. Since free neutrons are electrically neutral, they pass through the electrical fields within atoms without any interaction and they are interacting with matter almost exclusively through relatively rare collisions with atomic nuclei.

See also: Free Neutron

Shielding of Neutron Radiation
In radiation protection there are three ways how to protect people from identified radiation sources:
  • Limiting Time. The amount of radiation exposure depends directly (linearly) on the time people spend near the source of radiation. The dose can be reduced by limiting exposure time.
  • Distance. The amount of radiation exposure depends on the distance from the source of radiation. Similarly to a heat from a fire, if you are too close, the intensity of heat radiation is high and you can get burned. If you are at the right distance, you can withstand there without any problems and moreover it is comfortable. If you are too far from heat source, the insufficiency of heat can also hurt you. This analogy, in a certain sense, can be applied to radiation also from nuclear sources.
  • Shielding. Finally, if the source is too intensive and time or distance do not provide sufficient radiation protection the shielding must be used. Radiation shielding usually consist of barriers of lead, concrete or water. Even depleted uranium can be used as a good protection from gamma radiation, but on the other hand uranium is absolutely inappropriate shielding of neutron radiation. In short, it depends on type of radiation to be shielded, which shielding will be effective or not.

Shielding of Neutrons

Shielding of Neutron Radiation
Water as a neutron shield

There are three main features of neutrons, which are crucial in the shielding of neutrons.

  • Neutrons have no net electric charge, therefore they cannot be affected or stopped by electric forces. Neutrons ionize matter only indirectly, which makes neutrons highly penetrating type of radiation.
  • Neutrons scatter with heavy nuclei very elastically. Heavy nuclei very hard slow down a neutron let alone absorb a fast neutron.
  • An absorption of neutron (one would say shielding) causes initiation of certain nuclear reaction (e.g. radiative capture or even fission), which is accompanied by a number of other types of radiation. In short, neutrons make matter radioactive, therefore with neutrons we have to shield also the other types of radiation.

The best materials for shielding neutrons must be able to:

  • Slow down neutrons (the same principle as the neutron moderation). First point can be fulfilled only by material containing light atoms (e.g. hydrogen atoms), such as water, polyethylene, and concrete. The nucleus of a hydrogen nucleus contains only a proton. Since a proton and a neutron have almost identical masses, a neutron scattering on a hydrogen nucleus can give up a great amount of its energy (even entire kinetic energy of a neutron can be transferred to a proton after one collision). This is similar to a billiard. Since a cue ball and another billiard ball have identical masses, the cue ball hitting another ball can be made to stop and the other ball will start moving with the same velocity. On the other hand, if a ping pong ball is thrown against a bowling ball (neutron vs. heavy nucleus), the ping pong ball will bounce off with very little change in velocity, only a change in direction. Therefore lead is quite ineffective for blocking neutron radiation, as neutrons are uncharged and can simply pass through dense materials.
  • Table of cross-sections
    Table of cross-sections

    Absorb this slow neutron. Thermal neutrons can be easily absorbed by capture in materials with high neutron capture cross sections (thousands of barns) like boron, lithium or cadmium. Generally, only a thin layer of such absorbator is sufficient to shield thermal neutrons. Hydrogen (in the form of water), which can be used to slow down neutrons, have absorbtion cross-section 0.3 barns. This is not enough, but this insufficiency can be offset by sufficient thickness of water shield.

  • Shield the accompanying radiation. In the case of cadmium shield the absorption of neutrons is accompanied by strong emission of gamma rays. Therefore additional shield is necessary to attenuate the gamma rays. This phenomenon practically does not exist for lithium and is much less important for boron as a neutron absorption material. For this reason, materials containing boron are used often in neutron shields. In addition, boron (in the form of boric acid) is well soluble in water making this combination very efective neutron shield.

Water as a neutron shield

Water due to the high hydrogen content and the availability is efective and common neutron shielding. However, due to the low atomic number of hydrogen and oxygen, water is not acceptable shield against the gamma rays. On the other hand in some cases this disadvantage (low density) can be compensated by high thickness of the water shield.  In case of neutrons, water perfectly moderates neutrons, but with absorption of neutrons by hydrogen nucleus secondary gamma rays with the high energy are produced. These gamma rays highly penetrates matter and therefore it can increase requirements on the thickness of the water shield. Adding a boric acid can help with this problem (neutron absorbtion on boron nuclei without strong gamma emission), but results in another problems with corrosion of construction materials.

Concrete as a neutron shield

Most commonly used neutron shielding in many sectors of the nuclear science and engineering is shield of concrete. Concrete is also hydrogen-containing material, but unlike water concrete have higher density (suitable for secondary gamma shielding) and does not need any maintenance. Because concrete is a mixture of several different materials its composition is not constant. So when referring to concrete as a neutron shielding material, the material used in its composition should be told correctly. Generally concrete are divided to “ordinary “ concrete and “heavy” concrete. Heavy concrete uses heavy natural aggregates such as barites  (barium sulfate) or magnetite or manufactured aggregates such as iron, steel balls, steel punch or other additives. As a result of these additives, heavy concrete have higher density than ordinary concrete (~2300 kg/m3). Very heavy concrete can achieve density up to 5,900 kg/m3 with iron additives or up to 8900 kg/m3 with lead additives. Heavy concrete provide very effective protection against neutrons.

See also: Shielding of Neutron Radiation

Neutron Sources

A neutron source is any device that emits neutrons. Neutron sources have many applications, they can be used in research, engineering, medicine, petroleum exploration, biology, chemistry and nuclear power. A neutron source is characterized by a number of factors:

  • Significance of the source
  • Intensity. The rate of neutrons emitted by the source.
  • Energy distribution of emitted neutrons.
  • Angular distribution of emitted neutrons.
  • Mode of emission. Continuous or pulsed operation.

Classification by significance of the source

  • Large (Significant) neutron sources
    • Nuclear Reactors. There are nuclei that can undergo fission on their own spontaneously, but only certain nuclei, like uranium-235, uranium-233 and plutonium-239, can sustain a fission chain reaction. This is because these nuclei release neutrons when they break apart, and these neutrons can induce fission of other nuclei. Uranium-235 which exists as 0.7% of naturally occurring uranium undergoes nuclear fission with thermal neutrons with the production of, on average, 2.4 fast neutrons and the release of ~ 180 MeV of energy per fission. Free neutrons released by each fission play very important role as a trigger of the reaction, but they can be also used fo another purpose. For example: One neutron is required to trigger a further fission. Part of free neutrons (let say 0.5 neutrons/fission) is absorbed in other material, but an excess of neutrons (0.9 neutrons/fission) is able to leave the surface of the reactor core and can be used as a neutron source.
    • Fusion Systems. Nuclear fusion is a nuclear reaction in which two or more atomic nuclei (e.g. D+T) collide at a very high energy and fuse together. Thy byproduct of DT fusion is a free neutron (see picture), therefore also nuclear fusion reaction has the potential to produces large quantities of neutrons.
    • Spallation Sources. A spallation source is a high-flux neutron source in which protons that have been accelerated to high energies hit a heavy target material, causing the emission of neutrons. The reaction occurs above a certain energy threshold for the incident particle, which is typically 5 – 15 MeV.
  • Medium neutron sources
    • Bremssstrahlung from Electron Accelerators / Photofission. Energetic electrons when slowed down rapidly in a heavy target emit intense gamma radiation during the deceleration process. This is known as Bremsstrahlung or braking radiation. The interaction of the gamma radiation with the target produces neutrons via the (γ,n) reaction, or the (γ,fission) reaction when a fissile target is used. e-→Pb → γ→ Pb →(γ,n) and (γ,fission). The Bremsstrahlung γ energy exceeds the binding energy of the “last” neutron in the target. A source strength of 1013 neutrons/second produced in short (i.e. < 5 μs) pulses can be readily realised.
    • Dense plasme focus. The dense plasma focus (DPF) is a device that is known as an efficient source of neutrons from fusion reactions. Mechanism of dense plasma focus (DPF) is based on nuclear fusion of short-lived plasma of deuterium and/or tritium. This device produces a short-lived plasma by electromagnetic compression and acceleration that is called a pinch. This plasma is during the pinch hot and dense enough to cause nuclear fusion and the emission of neutrons.
    • Light ion accelerators. Neutrons can be also produced by particle accelerators using targets of deuterium, tritium, lithium, beryllium, and other low-Z materials. In this case the target must be bombarded with accelerated hydrogen (H), deuterium (D), or tritium (T) nuclei.
  • Small neutron sources
    • Neutron Generators. Neutrons are produced in the fusion of deuterium and tritium in the following exothermic reaction. 2D + 3T → 4He + n + 17.6 MeV.  The neutron is produced with a kinetic energy of 14.1 MeV. This can be achieved on a small scale in the laboratory with a modest 100 kV accelerator for deuterium atoms bombarding a tritium target. Continuous neutron sources of ~1011 neutrons/second can be achieved relatively simply.
    • Radioisotope source – (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. 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.
    • Radioisotope source – (γ,n) reactions. (γ,n) reactions can also be used for the same purpose. In this type of source, because of the greater range of the γ-ray, the two physical  components of the source can be separated making it possible to ‘switch off’ the reaction if so required by removing the radioactive source from the beryllium. (γ,n) sources produce a monoenergetic neutrons unlike (α,n) sources.  The (γ,n) source uses antimony-124 as the gamma emitter in the following endothermic reaction.

124Sb→124Te + β− + γ

γ + 9Be→8Be + n – 1.66 MeV

    • Radioisotope source – spontaneous fission. Certain isotopes undergo spontaneous fission with emission of neutrons. The most commonly used spontaneous fission source is the radioactive isotope californium-252. Cf-252 and all 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.

See also: Neutron Sources

See also: Source Neutrons

Prompt and Delayed Neutrons
It is known the fission neutrons are of importance in any chain-reacting system. Neutrons trigger the nuclear fission of some nuclei (235U, 238U or even 232Th). What is crucial the fission of such nuclei produces 2, 3 or more free neutrons.

But not all neutrons are released at the same time following fission. Even the nature of creation of these neutrons is different. From this point of view we usually divide the fission neutrons into two following groups:

  • Prompt Neutrons. Prompt neutrons are emitted directly from fission and they are emitted within very short time of about 10-14 second.
  • Delayed Neutrons. Delayed neutrons are emitted by neutron rich fission fragments that are called the delayed neutron precursors. These precursors usually undergo beta decay but a small fraction of them are excited enough to undergo neutron emission. The fact the neutron is produced via this type of decay and this happens orders of magnitude later compared to the emission of the prompt neutrons, plays an extremely important role in the control of the reactor.

Table of key prompt and delayed neutrons characteristics

What is Shielding of Spent Fuel in Spent Fuel Pool – Definition

As was written, shielding of the spent fuel pool is ensured by: the design of the spent fuel pool (water and concrete shielding), the design of surrounding working area, requirements on water level in the pool. Reactor Physics
Spent fuel pool
Spent fuel pool. Source: wikipedia.org License: Public Domain

Spent fuel pool (SFP) is storage pool for spent nuclear fuel from nuclear reactors. Spent fuel pool may be located inside the containment building or inside the fuel building (outside the containment building). When located outside the containment building, the two areas are connected by a fuel transfer system which carries the fuel through a normally closed opening in the reactor containment. In this case spent fuel is removed from the reactor vessel by a manipulator crane and placed in the fuel transfer system. In the spent fuel pool, the fuel is removed from the transfer system and placed into storage racks. After a suitable decay period, the fuel can be removed from storage and loaded into a shipping cask for removal from the site. Spent fuel pools are typically 12m or more deep, with the bottom equipped with storage racks designed to hold fuel assemblies removed from the reactor. A reactor’s pool is specially designed for the reactor in which the fuel was used and situated at the reactor site.

Shielding of Spent Fuel in Spent Fuel Pool

It must be noted, irradiated fuel is due to presence of high amount of radioactive fission fragments and transuranic elements very hot and very radioactive. Reactor operators have to manage the heat and radioactivity that remains in the “spent fuel” after it’s taken out of the reactor.  In nuclear power plants, spent nuclear fuel is usually stored underwater in the spent fuel pool on the plant. Plant personnel move the spent fuel underwater from the reactor to the pool. Over time, as the spent fuel is stored in the pool, it becomes cooler as the radioactivity decays away. After several years (> 5 years), decay heat and radioactivity decreases under specified limits so that spent fuel may be reprocessed or interim storaged.

As was written, shielding of the spent fuel pool is ensured by:

  • the design of the spent fuel pool (water and concrete shielding),
  • the design of surrounding working area
  • requirements on water level in the pool,

Spent fuel pools are fitted with stainless steel and aluminum racks that hold the fuel assemblies and are lined with stainless steel to prevent leaking. There are no drains that would allow the water level to drop or the pool to become empty. The plants have a variety of extra water sources and equipment to replenish water that evaporates over time, or in case there is a leak. Plant personnel are also trained and prepared to quickly respond to a problem. The water serves two purposes: it cools the fuel and shields workers at the plant from radioactivity. Although water is neither high density nor high Z material, it is commonly used as gamma shields. Water provides a radiation shielding of fuel assemblies in a spent fuel pool during storage or during transports from and into the reactor core. Although water is a low-density material and low Z material, it is commonly used in nuclear power plants, because these disadvantages can be compensated with increased thickness.

The minimum water level in the fuel storage pool meets primarily the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel. But it also provides with the coolant required for cooling spent fuel.

Spent Fuel – Cherenkov Radiation

Cherenkov Radiation in the reactor core.
Cherenkov Radiation in the reactor core.

The cherenkov radiation is electromagnetic radiation emitted when a charged particle (such as an electron) moves through a dielectric medium faster than the phase velocity of light in that medium.

Cherenkov radiation can be used to detect high-energy charged particles (especially beta particles). In nuclear reactors or in a spent nuclear fuel pool, beta particles (high-energy electrons) are released as the fission fragments decay. The glow is visible also after the chain reaction stops (in the reactor). The cherenkov radiation can characterize the remaining radioactivity of spent nuclear fuel, therefore it can be used for measuring of fuel burnup.

References:
Nuclear and Reactor Physics:
  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
  3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2.
  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See also:

Spent Fuel Pool

We hope, this article, Shielding of Spent Fuel in Spent Fuel Pool, helps you. If so, give us a like in the sidebar. Main purpose of this website is to help the public to learn some interesting and important information about physics and reactor physics.

What is Cooling of Spent Fuel Pool – Definition

As was written, cooling of the spent fuel pool is ensured by: the design of the spent fuel pool (e.g. no drains below the top of the stored fuel elements), requirements on water level in the pool, requirements on active cooling elements (heat exchangers and heat sink). Reactor Physics
Spent fuel pool
Spent fuel pool. Source: wikipedia.org License: Public Domain

Spent fuel pool (SFP) is storage pool for spent nuclear fuel from nuclear reactors. Spent fuel pool may be located inside the containment building or inside the fuel building (outside the containment building). When located outside the containment building, the two areas are connected by a fuel transfer system which carries the fuel through a normally closed opening in the reactor containment. In this case spent fuel is removed from the reactor vessel by a manipulator crane and placed in the fuel transfer system. In the spent fuel pool, the fuel is removed from the transfer system and placed into storage racks. After a suitable decay period, the fuel can be removed from storage and loaded into a shipping cask for removal from the site. Spent fuel pools are typically 12m or more deep, with the bottom equipped with storage racks designed to hold fuel assemblies removed from the reactor. A reactor’s pool is specially designed for the reactor in which the fuel was used and situated at the reactor site.

Cooling of Spent Fuel Pool

Decay HeatIt must be noted, irradiated fuel is due to presence of high amount of radioactive fission fragments and transuranic elements very hot and very radioactive. Reactor operators have to manage the heat and radioactivity that remains in the “spent fuel” after it’s taken out of the reactor.  In nuclear power plants, spent nuclear fuel is usually stored underwater in the spent fuel pool on the plant. Plant personnel move the spent fuel underwater from the reactor to the pool. Over time, as the spent fuel is stored in the pool, it becomes cooler as the radioactivity decays away. After several years (> 5 years), decay heat decreases under specified limits so that spent fuel may be reprocessed or interim storaged.

However, even at these low levels (about 0.25% after ten days), the amount of heat generated requires the continued removal of heat for an appreciable time after shutdown. Decay heat is a long-term consideration and impacts spent fuel handling, reprocessing, waste management, and reactor safety.

The design of the reactor and the spent fuel pool must allow for the removal of this decay heat from the core by some means. If adequate heat removal is not available, decay heat will increase the temperatures in the fuel to the point that fuel melting and fuel assembly damage will occur as in the case of Fukushima. The degree of concern with decay heat will vary according to reactor type and design.

As was written, cooling of the spent fuel pool is ensured by:

  • the design of the spent fuel pool (e.g. no drains below the top of the stored fuel elements),
  • requirements on water level in the pool,
  • requirements on active cooling elements (heat exchangers and heat sink).

The minimum water level in the fuel storage pool meets primarily the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel. But it also provides with the coolant required for cooling spent fuel.

Cooling of the spent fuel pool is usually ensured by the spent fuel pool cooling system, which is designed to remove residual decay heat generated by spent fuel stored in the spent fuel pool. The system also maintains the purity of the spent fuel cooling water and the refueling water. The system is designed to provide with cooling of stored spent fuel and it may become necessary to cool all fuel assemblies from totally unloaded reactor. The design incorporates redundant active components (usually 3×100%) and it is one of safety systems. One of principal safety functions of the ultimate heat sink (UHS) is dissipation of maximum expected decay heat from the spent fuel pool to ensure the pool temperature remains within the design bounds for the structure.

References:
Nuclear and Reactor Physics:
  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
  3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2.
  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See also:

Spent Fuel Pool

We hope, this article, Cooling of Spent Fuel Pool, helps you. If so, give us a like in the sidebar. Main purpose of this website is to help the public to learn some interesting and important information about physics and reactor physics.

What is Burnup Credit in Nuclear Criticality Analysis – Definition

In criticality licensing of spent fuel pool, taking credit for the decrease in fuel reactivity due to fuel burnup is known as burnup credit. Reactor Physics
Spent fuel pool
Spent fuel pool. Source: wikipedia.org License: Public Domain

As was written, subcriticality of the spent fuel pool is ensured by:

  • the design of the spent fuel pool,
  • requirements on boric acid diluted in water,
  • limiting of stored fuel (e.g. fuel enrichment, assembly burnup)

Today, spent fuel is usually stored in so called high-density racks or in the maximum-density rack (MDR).  Using such racks, fuel assemblies can be stored in about one half the volume required for storage in standard racks. Higher storage densities have been achieved without the risk of a nuclear chain reaction by adding neutron absorbing materials (typically boron) in storage racks and baskets, and dissolved in the water itself. These racks incorporate (boron-10) or other neutron-absorbing material to ensure subcriticality. Boron-10 is generally present in the chemical form of boron carbide (B4C) within a metal matrix (e.g., Boral and Metamic (trademark of Metamic, LLC)) or a polymer matrix (e.g., Boraflex (trademark of BISCO), Carborundum, and Tetrabor), although borated stainless steel incorporates the boron-10 atoms into the alloy composition.

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. In Region 2, maximum reactivity fuel is not allowed to load, while it can be loaded in the racks of Region 1 of the pool.

Most conservative approach requires that the multiplication factor, assuming flooding with pure water and infinite geometry, does not exceed 0.95 with a full loading of the maximum anticipated enrichment. To satisfy this design criterion, the assumptions in the criticality evaluation are as below:

  • The fuel assemblies have the maximum approved initial enrichment with the highest reactivity in fuel’s lifetime, and without the control rods (burnable poisons may be taken into account).
  • In the flooding condition, all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 (5% subcriticality) be evaluated in the absence of soluble boron.
  • The array of the fuel assemblies can be taken as infinite geometry, or with reflective boundary condition.
  • The effect of structure material and the fixed neutron absorber can be considered.
  • Unless the double contingency principle is taken, the presence of the boron in the moderation should not be considered. This principle shows at least two independent, unlikely and concurrent incidents have to happen to lead a criticality accident.

Burnup Credit

In criticality licensing of spent fuel pool, taking credit for the decrease in fuel reactivity due to fuel burnup is known as burnup credit. Burnup credit is similar to boron credit. The concept of burnup credit is taking credit for the reduction in reactivity due to irradiation of nuclear fuel when the criticality safety analysis is carried out for the spent fuel. In spent fuel storage, it has generally been required that for criticality analyses that it be assumed that fuel si at its peak reactivity, which is generally the fresh fuel. But in the spent fuel pool, most of fuel assemblies have higher burnup with lower reactivity.

The reduction of reactivity is a combinative effect of:

  • the net reduction of fissile nuclides,
  • the production of neutron-absorbing nuclides (non-fissile actinides and fission products)

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. There are two kinds of storage racks:

  • The storage racks applied to the spent fuel assemblies for which the burnup values must be equal or more than the burnup value credited in the criticality safety calculation. (Region 2)
  • The storage racks applied to the spent fuel assemblies assuming non-irradiated, with the maximum reactivity (Region 1).

It should be clear that the burnup credit is not attempt to reduce the safety margins in criticality safety. It is just to reduce the analysis conservatism, in another word, reduce the uncertainties in safety margins by a more accurate safety analysis. The fresh fuel assumption can be very conservative and result in a significant reduction in capacity for a given storage or cask volume. The issue is complicated by ensuring that adequate administrative controls and measurement systems exist to prevent higher reactivity fuel from being placed in storage racks. Regulators also have concerns about verifying the accuracy of being allowed burnup credit.

For example, abnormal conditions should include consideration of fuel assemblies loaded into storage racks not approved for their storage (e.g., fuel not meeting minimum burnup requirements stored in burnup credit storage racks).

Special reference: Introduction of Burn-up Credit in Nuclear Criticality Safety Analysis. Guoshun You, Chunming Zhang, Xinyi Pan. Nuclear and Radiation Safety Center of MEP.

&nbsp;

References:
Nuclear and Reactor Physics:
  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
  3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2.
  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See also:

Spent Fuel Pool

We hope, this article, Burnup Credit in Nuclear Criticality Analysis, helps you. If so, give us a like in the sidebar. Main purpose of this website is to help the public to learn some interesting and important information about physics and reactor physics.

What is Double Contingency Principle – Definition

The double contingency principle discussed in ANSI/ANS-8.1 allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. Reactor Physics
Spent fuel pool
Spent fuel pool. Source: wikipedia.org License: Public Domain

“The double contingency approach requires a demonstration that unintended criticality cannot occur unless at least two unlikely, independent, concurrent changes in the conditions originally specified as essential to criticality safety have occurred.”

Source: Nuclear Safety Technical Assessment Guide. NS-TAST-GD-041 Revision 5. ONR, 2016.

The double contingency principle discussed in ANSI/ANS-8.1 allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the movement of fuel within spent fuel pool, and accidental misloading of a fuel assembly in the Region 2 of the spent fuel pool. This could potentially increase the criticality of Region 2. To mitigate these postulated criticality related accidents, boron is dissolved in the pool water.

References:
Nuclear and Reactor Physics:
  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
  3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2.
  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See also:

Spent Fuel Pool

We hope, this article, Double Contingency Principle, helps you. If so, give us a like in the sidebar. Main purpose of this website is to help the public to learn some interesting and important information about physics and reactor physics.

What is Boron Credit – Partial Boron Credit – Definition

The credit for soluble boron (boron credit or partial boron credit – PBC) in the spent fuel pool criticality analysis offers a solution to these concerns. Reactor Physics
Spent fuel pool
Spent fuel pool. Source: wikipedia.org License: Public Domain

As was written, subcriticality of the spent fuel pool is ensured by:

  • the design of the spent fuel pool,
  • requirements on boric acid diluted in water,
  • limiting of stored fuel (e.g. fuel enrichment, assembly burnup)

Today, spent fuel is usually stored in so called high-density racks or in the maximum-density rack (MDR).  Using such racks, fuel assemblies can be stored in about one half the volume required for storage in standard racks. Higher storage densities have been achieved without the risk of a nuclear chain reaction by adding neutron absorbing materials (typically boron) in storage racks and baskets, and dissolved in the water itself. These racks incorporate (boron-10) or other neutron-absorbing material to ensure subcriticality. Boron-10 is generally present in the chemical form of boron carbide (B4C) within a metal matrix (e.g., Boral and Metamic (trademark of Metamic, LLC)) or a polymer matrix (e.g., Boraflex (trademark of BISCO), Carborundum, and Tetrabor), although borated stainless steel incorporates the boron-10 atoms into the alloy composition.

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. In Region 2, maximum reactivity fuel is not allowed to load, while it can be loaded in the racks of Region 1 of the pool.

Most conservative approach requires that the multiplication factor, assuming flooding with pure water and infinite geometry, does not exceed 0.95 with a full loading of the maximum anticipated enrichment. To satisfy this design criterion, the assumptions in the criticality evaluation are as below:

  • The fuel assemblies have the maximum approved initial enrichment with the highest reactivity in fuel’s lifetime, and without the control rods (burnable poisons may be taken into account).
  • In the flooding condition, all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 (5% subcriticality) be evaluated in the absence of soluble boron.
  • The array of the fuel assemblies can be taken as infinite geometry, or with reflective boundary condition.
  • The effect of structure material and the fixed neutron absorber can be considered.
  • Unless the double contingency principle is taken, the presence of the boron in the moderation should not be considered. This principle shows at least two independent, unlikely and concurrent incidents have to happen to lead a criticality accident.

Boron Credit

The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. We mean that boric acid is dissolved in the coolant. Boric acid (molecular formula: H3BO3), is a white powder that is soluble in water. According to the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 (5% subcriticality) be evaluated in the absence of soluble boron. Hence, the design of the spent fuel pool is based on the use of unborated water. Unless the double contingency principle is taken, the presence of the boron (boron credit) in the moderation should not be considered. Nuclear plant owners are facing increasing fuel assembly enrichments, spent-fuel assembly burnup limitations, spent fuel pool storage cell restrictions, and problems with fixed neutron absorber degradation, all of which are challenging their traditional criticality analyses. The credit for soluble boron (boron credit or partial boron credit – PBC) in the spent fuel pool criticality analysis offers a solution to these concerns.

For example, according to NUREG-0800 (9.1.1-4):

“For PWR pools where partial credit for soluble boron is taken, both of the following criteria must be met:

  1. When the spent fuel storage racks are loaded with fuel of the maximum permissible reactivity and are flooded with full-density unborated water, the maximum K(eff) must be less than 1.0 for all normal and credible abnormal conditions. The K(eff) must include allowance for all relevant uncertainties and tolerances.
  2. When the spent fuel storage racks are loaded with fuel of the maximum permissible reactivity and are flooded with full-density water borated to a minimum concentration (CB,min, measured in parts per million of boron), the maximum K(eff) must be no greater than 0.95 for all normal conditions. Plant technical specifications must incorporate the CB,min. The K(eff) must include allowance for all relevant uncertainties and tolerances.”

Double Contingency Principle – Double Contingency Approach

“The double contingency approach requires a demonstration that unintended criticality cannot occur unless at least two unlikely, independent, concurrent changes in the conditions originally specified as essential to criticality safety have occurred.”

Source: Nuclear Safety Technical Assessment Guide. NS-TAST-GD-041 Revision 5. ONR, 2016.

The double contingency principle discussed in ANSI/ANS-8.1 allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the movement of fuel within spent fuel pool, and accidental misloading of a fuel assembly in the Region 2 of the spent fuel pool. This could potentially increase the criticality of Region 2. To mitigate these postulated criticality related accidents, boron is dissolved in the pool water.

References:
Nuclear and Reactor Physics:
  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
  3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2.
  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See also:

Spent Fuel Pool

We hope, this article, Boron Credit – Partial Boron Credit, helps you. If so, give us a like in the sidebar. Main purpose of this website is to help the public to learn some interesting and important information about physics and reactor physics.

What is High Density Rack – Maximum Density Rack – Definition

Today, spent fuel is usually stored in so called high density racks or in the maximum density rack (MDR).  Using such racks, fuel assemblies can be stored in about one half the volume required for storage in standard racks. Reactor Physics
Spent fuel pool
Spent fuel pool. Source: wikipedia.org License: Public Domain

Spent fuel pool (SFP) is storage pool for spent nuclear fuel from nuclear reactors. Spent fuel pool may be located inside the containment building or inside the fuel building (outside the containment building). When located outside the containment building, the two areas are connected by a fuel transfer system which carries the fuel through a normally closed opening in the reactor containment. In this case spent fuel is removed from the reactor vessel by a manipulator crane and placed in the fuel transfer system. In the spent fuel pool, the fuel is removed from the transfer system and placed into storage racks. After a suitable decay period, the fuel can be removed from storage and loaded into a shipping cask for removal from the site. Spent fuel pools are typically 12m or more deep, with the bottom equipped with storage racks designed to hold fuel assemblies removed from the reactor. A reactor’s pool is specially designed for the reactor in which the fuel was used and situated at the reactor site.

High Density Racks

As was written, subcriticality of the spent fuel pool is ensured by:

  • the design of the spent fuel pool,
  • requirements on boric acid diluted in water,
  • limiting of stored fuel (e.g. fuel enrichment, assembly burnup)

Today, spent fuel is usually stored in so called high-density racks or in the maximum-density rack (MDR).  Using such racks, fuel assemblies can be stored in about one half the volume required for storage in standard racks. Higher storage densities have been achieved without the risk of a nuclear chain reaction by adding neutron absorbing materials (typically boron) in storage racks and baskets, and dissolved in the water itself. These racks incorporate (boron-10) or other neutron-absorbing material to ensure subcriticality. Boron-10 is generally present in the chemical form of boron carbide (B4C) within a metal matrix (e.g., Boral and Metamic (trademark of Metamic, LLC)) or a polymer matrix (e.g., Boraflex (trademark of BISCO), Carborundum, and Tetrabor), although borated stainless steel incorporates the boron-10 atoms into the alloy composition.

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. In Region 2, maximum reactivity fuel is not allowed to load, while it can be loaded in the racks of Region 1 of the pool.

Most conservative approach requires that the multiplication factor, assuming flooding with pure water and infinite geometry, does not exceed 0.95 with a full loading of the maximum anticipated enrichment. To satisfy this design criterion, the assumptions in the criticality evaluation are as below:

  • The fuel assemblies have the maximum approved initial enrichment with the highest reactivity in fuel’s lifetime, and without the control rods (burnable poisons may be taken into account).
  • In the flooding condition, all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 (5% subcriticality) be evaluated in the absence of soluble boron.
  • The array of the fuel assemblies can be taken as infinite geometry, or with reflective boundary condition.
  • The effect of structure material and the fixed neutron absorber can be considered.
  • Unless the double contingency principle is taken, the presence of the boron in the moderation should not be considered. This principle shows at least two independent, unlikely and concurrent incidents have to happen to lead a criticality accident.

Burnup Credit

In criticality licensing of spent fuel pool, taking credit for the decrease in fuel reactivity due to fuel burnup is known as burnup credit. Burnup credit is similar to boron credit. The concept of burnup credit is taking credit for the reduction in reactivity due to irradiation of nuclear fuel when the criticality safety analysis is carried out for the spent fuel. In spent fuel storage, it has generally been required that for criticality analyses that it be assumed that fuel si at its peak reactivity, which is generally the fresh fuel. But in the spent fuel pool, most of fuel assemblies have higher burnup with lower reactivity.

The reduction of reactivity is a combinative effect of:

  • the net reduction of fissile nuclides,
  • the production of neutron-absorbing nuclides (non-fissile actinides and fission products)

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. There are two kinds of storage racks:

  • The storage racks applied to the spent fuel assemblies for which the burnup values must be equal or more than the burnup value credited in the criticality safety calculation. (Region 2)
  • The storage racks applied to the spent fuel assemblies assuming non-irradiated, with the maximum reactivity (Region 1).

It should be clear that the burnup credit is not attempt to reduce the safety margins in criticality safety. It is just to reduce the analysis conservatism, in another word, reduce the uncertainties in safety margins by a more accurate safety analysis. The fresh fuel assumption can be very conservative and result in a significant reduction in capacity for a given storage or cask volume. The issue is complicated by ensuring that adequate administrative controls and measurement systems exist to prevent higher reactivity fuel from being placed in storage racks. Regulators also have concerns about verifying the accuracy of being allowed burnup credit.

For example, abnormal conditions should include consideration of fuel assemblies loaded into storage racks not approved for their storage (e.g., fuel not meeting minimum burnup requirements stored in burnup credit storage racks).

Special reference: Introduction of Burn-up Credit in Nuclear Criticality Safety Analysis. Guoshun You, Chunming Zhang, Xinyi Pan. Nuclear and Radiation Safety Center of MEP.

References:
Nuclear and Reactor Physics:
  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
  3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2.
  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See also:

Spent Fuel Pool

We hope, this article, High Density Rack – Maximum Density Rack, helps you. If so, give us a like in the sidebar. Main purpose of this website is to help the public to learn some interesting and important information about physics and reactor physics.

What is Subcriticality of Spent Fuel Pool – Definition

As was written, subcriticality of the spent fuel pool is ensured by: the design of the spent fuel pool, requirements on boric acid diluted in water, limiting of stored fuel
Spent fuel pool
Spent fuel pool. Source: wikipedia.org License: Public Domain

As was written, subcriticality of the spent fuel pool is ensured by:

  • the design of the spent fuel pool,
  • requirements on boric acid diluted in water,
  • limiting of stored fuel (e.g. fuel enrichment, assembly burnup)

Today, spent fuel is usually stored in so called high-density racks or in the maximum-density rack (MDR).  Using such racks, fuel assemblies can be stored in about one half the volume required for storage in standard racks. Higher storage densities have been achieved without the risk of a nuclear chain reaction by adding neutron absorbing materials (typically boron) in storage racks and baskets, and dissolved in the water itself. These racks incorporate (boron-10) or other neutron-absorbing material to ensure subcriticality. Boron-10 is generally present in the chemical form of boron carbide (B4C) within a metal matrix (e.g., Boral and Metamic (trademark of Metamic, LLC)) or a polymer matrix (e.g., Boraflex (trademark of BISCO), Carborundum, and Tetrabor), although borated stainless steel incorporates the boron-10 atoms into the alloy composition.

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. In Region 2, maximum reactivity fuel is not allowed to load, while it can be loaded in the racks of Region 1 of the pool.

Most conservative approach requires that the multiplication factor, assuming flooding with pure water and infinite geometry, does not exceed 0.95 with a full loading of the maximum anticipated enrichment. To satisfy this design criterion, the assumptions in the criticality evaluation are as below:

  • The fuel assemblies have the maximum approved initial enrichment with the highest reactivity in fuel’s lifetime, and without the control rods (burnable poisons may be taken into account).
  • In the flooding condition, all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 (5% subcriticality) be evaluated in the absence of soluble boron.
  • The array of the fuel assemblies can be taken as infinite geometry, or with reflective boundary condition.
  • The effect of structure material and the fixed neutron absorber can be considered.
  • Unless the double contingency principle is taken, the presence of the boron in the moderation should not be considered. This principle shows at least two independent, unlikely and concurrent incidents have to happen to lead a criticality accident.

Boron Credit

The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. We mean that boric acid is dissolved in the coolant. Boric acid (molecular formula: H3BO3), is a white powder that is soluble in water. According to the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 (5% subcriticality) be evaluated in the absence of soluble boron. Hence, the design of the spent fuel pool is based on the use of unborated water. Unless the double contingency principle is taken, the presence of the boron (boron credit) in the moderation should not be considered. Nuclear plant owners are facing increasing fuel assembly enrichments, spent-fuel assembly burnup limitations, spent fuel pool storage cell restrictions, and problems with fixed neutron absorber degradation, all of which are challenging their traditional criticality analyses. The credit for soluble boron (boron credit or partial boron credit – PBC) in the spent fuel pool criticality analysis offers a solution to these concerns.

For example, according to NUREG-0800 (9.1.1-4):

“For PWR pools where partial credit for soluble boron is taken, both of the following criteria must be met:

  1. When the spent fuel storage racks are loaded with fuel of the maximum permissible reactivity and are flooded with full-density unborated water, the maximum K(eff) must be less than 1.0 for all normal and credible abnormal conditions. The K(eff) must include allowance for all relevant uncertainties and tolerances.
  2. When the spent fuel storage racks are loaded with fuel of the maximum permissible reactivity and are flooded with full-density water borated to a minimum concentration (CB,min, measured in parts per million of boron), the maximum K(eff) must be no greater than 0.95 for all normal conditions. Plant technical specifications must incorporate the CB,min. The K(eff) must include allowance for all relevant uncertainties and tolerances.”

Double Contingency Principle – Double Contingency Approach

“The double contingency approach requires a demonstration that unintended criticality cannot occur unless at least two unlikely, independent, concurrent changes in the conditions originally specified as essential to criticality safety have occurred.”

Source: Nuclear Safety Technical Assessment Guide. NS-TAST-GD-041 Revision 5. ONR, 2016.

The double contingency principle discussed in ANSI/ANS-8.1 allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the movement of fuel within spent fuel pool, and accidental misloading of a fuel assembly in the Region 2 of the spent fuel pool. This could potentially increase the criticality of Region 2. To mitigate these postulated criticality related accidents, boron is dissolved in the pool water.

Burnup Credit

In criticality licensing of spent fuel pool, taking credit for the decrease in fuel reactivity due to fuel burnup is known as burnup credit. Burnup credit is similar to boron credit. The concept of burnup credit is taking credit for the reduction in reactivity due to irradiation of nuclear fuel when the criticality safety analysis is carried out for the spent fuel. In spent fuel storage, it has generally been required that for criticality analyses that it be assumed that fuel si at its peak reactivity, which is generally the fresh fuel. But in the spent fuel pool, most of fuel assemblies have higher burnup with lower reactivity.

The reduction of reactivity is a combinative effect of:

  • the net reduction of fissile nuclides,
  • the production of neutron-absorbing nuclides (non-fissile actinides and fission products)

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. There are two kinds of storage racks:

  • The storage racks applied to the spent fuel assemblies for which the burnup values must be equal or more than the burnup value credited in the criticality safety calculation. (Region 2)
  • The storage racks applied to the spent fuel assemblies assuming non-irradiated, with the maximum reactivity (Region 1).

It should be clear that the burnup credit is not attempt to reduce the safety margins in criticality safety. It is just to reduce the analysis conservatism, in another word, reduce the uncertainties in safety margins by a more accurate safety analysis. The fresh fuel assumption can be very conservative and result in a significant reduction in capacity for a given storage or cask volume. The issue is complicated by ensuring that adequate administrative controls and measurement systems exist to prevent higher reactivity fuel from being placed in storage racks. Regulators also have concerns about verifying the accuracy of being allowed burnup credit.

For example, abnormal conditions should include consideration of fuel assemblies loaded into storage racks not approved for their storage (e.g., fuel not meeting minimum burnup requirements stored in burnup credit storage racks).

Special reference: Introduction of Burn-up Credit in Nuclear Criticality Safety Analysis. Guoshun You, Chunming Zhang, Xinyi Pan. Nuclear and Radiation Safety Center of MEP.

References:
Nuclear and Reactor Physics:
  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
  3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2.
  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See also:

Spent Fuel Pool

We hope, this article, Subcriticality of Spent Fuel Pool, helps you. If so, give us a like in the sidebar. Main purpose of this website is to help the public to learn some interesting and important information about physics and reactor physics.

What is Safety of Spent Fuel Pool – Definition

Safety of spent fuel pool stands on various criteria. These criteria may be grouped according to following aspects: subcriticality, cooling, shielding and integrity. Reactor Physics
Spent fuel pool
Spent fuel pool. Source: wikipedia.org License: Public Domain

Spent fuel pool (SFP) is storage pool for spent nuclear fuel from nuclear reactors. Spent fuel pool may be located inside the containment building or inside the fuel building (outside the containment building). When located outside the containment building, the two areas are connected by a fuel transfer system which carries the fuel through a normally closed opening in the reactor containment. In this case spent fuel is removed from the reactor vessel by a manipulator crane and placed in the fuel transfer system. In the spent fuel pool, the fuel is removed from the transfer system and placed into storage racks. After a suitable decay period, the fuel can be removed from storage and loaded into a shipping cask for removal from the site. Spent fuel pools are typically 12m or more deep, with the bottom equipped with storage racks designed to hold fuel assemblies removed from the reactor. A reactor’s pool is specially designed for the reactor in which the fuel was used and situated at the reactor site.

Safety of Spent Fuel Pool

Safety of spent fuel pools stands on various criteria. These criteria may be grouped according to following aspects:

  • Subcriticality. Fulfillment of this criterion is based on:
    • the design of the spent fuel pool,
    • requirements on boric acid diluted in water,
    • limiting of stored fuel (e.g. fuel enrichment, assembly burnup)
  • Cooling. Fulfillment of this criterion is based on:
    • the design of the spent fuel pool (e.g. no drains below the top of the stored fuel elements),
    • requirements on water level in the pool,
    • requirements on active cooling elements (heat exchangers and heat sink).
  • Radiation Shielding. Fulfillment of this criterion is based on:
    • the design of the spent fuel pool (water and concrete shielding),
    • the design of surrounding working area
    • requirements on water level in the pool,
  • Integrity. Fulfillment of this criterion is based on:
    • the design of the spent fuel pool,
    • the design of surrounding working area,
    • ensuring periodic inspections

Subcriticality of Spent Fuel Pool

As was written, subcriticality of the spent fuel pool is ensured by:

  • the design of the spent fuel pool,
  • requirements on boric acid diluted in water,
  • limiting of stored fuel (e.g. fuel enrichment, assembly burnup)

Today, spent fuel is usually stored in so called high-density racks or in the maximum-density rack (MDR).  Using such racks, fuel assemblies can be stored in about one half the volume required for storage in standard racks. Higher storage densities have been achieved without the risk of a nuclear chain reaction by adding neutron absorbing materials (typically boron) in storage racks and baskets, and dissolved in the water itself. These racks incorporate (boron-10) or other neutron-absorbing material to ensure subcriticality. Boron-10 is generally present in the chemical form of boron carbide (B4C) within a metal matrix (e.g., Boral and Metamic (trademark of Metamic, LLC)) or a polymer matrix (e.g., Boraflex (trademark of BISCO), Carborundum, and Tetrabor), although borated stainless steel incorporates the boron-10 atoms into the alloy composition.

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. In Region 2, maximum reactivity fuel is not allowed to load, while it can be loaded in the racks of Region 1 of the pool.

Most conservative approach requires that the multiplication factor, assuming flooding with pure water and infinite geometry, does not exceed 0.95 with a full loading of the maximum anticipated enrichment. To satisfy this design criterion, the assumptions in the criticality evaluation are as below:

  • The fuel assemblies have the maximum approved initial enrichment with the highest reactivity in fuel’s lifetime, and without the control rods (burnable poisons may be taken into account).
  • In the flooding condition, all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 (5% subcriticality) be evaluated in the absence of soluble boron.
  • The array of the fuel assemblies can be taken as infinite geometry, or with reflective boundary condition.
  • The effect of structure material and the fixed neutron absorber can be considered.
  • Unless the double contingency principle is taken, the presence of the boron in the moderation should not be considered. This principle shows at least two independent, unlikely and concurrent incidents have to happen to lead a criticality accident.

Cooling of Spent Fuel Pool

Decay HeatIt must be noted, irradiated fuel is due to presence of high amount of radioactive fission fragments and transuranic elements very hot and very radioactive. Reactor operators have to manage the heat and radioactivity that remains in the “spent fuel” after it’s taken out of the reactor.  In nuclear power plants, spent nuclear fuel is usually stored underwater in the spent fuel pool on the plant. Plant personnel move the spent fuel underwater from the reactor to the pool. Over time, as the spent fuel is stored in the pool, it becomes cooler as the radioactivity decays away. After several years (> 5 years), decay heat decreases under specified limits so that spent fuel may be reprocessed or interim storaged.

However, even at these low levels (about 0.25% after ten days), the amount of heat generated requires the continued removal of heat for an appreciable time after shutdown. Decay heat is a long-term consideration and impacts spent fuel handling, reprocessing, waste management, and reactor safety.

The design of the reactor and the spent fuel pool must allow for the removal of this decay heat from the core by some means. If adequate heat removal is not available, decay heat will increase the temperatures in the fuel to the point that fuel melting and fuel assembly damage will occur as in the case of Fukushima. The degree of concern with decay heat will vary according to reactor type and design.

As was written, cooling of the spent fuel pool is ensured by:

  • the design of the spent fuel pool (e.g. no drains below the top of the stored fuel elements),
  • requirements on water level in the pool,
  • requirements on active cooling elements (heat exchangers and heat sink).

The minimum water level in the fuel storage pool meets primarily the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel. But it also provides with the coolant required for cooling spent fuel.

Cooling of the spent fuel pool is usually ensured by the spent fuel pool cooling system, which is designed to remove residual decay heat generated by spent fuel stored in the spent fuel pool. The system also maintains the purity of the spent fuel cooling water and the refueling water. The system is designed to provide with cooling of stored spent fuel and it may become necessary to cool all fuel assemblies from totally unloaded reactor. The design incorporates redundant active components (usually 3×100%) and it is one of safety systems. One of principal safety functions of the ultimate heat sink (UHS) is dissipation of maximum expected decay heat from the spent fuel pool to ensure the pool temperature remains within the design bounds for the structure.

Shielding of Spent Fuel in Spent Fuel Pool

It must be noted, irradiated fuel is due to presence of high amount of radioactive fission fragments and transuranic elements very hot and very radioactive. Reactor operators have to manage the heat and radioactivity that remains in the “spent fuel” after it’s taken out of the reactor.  In nuclear power plants, spent nuclear fuel is usually stored underwater in the spent fuel pool on the plant. Plant personnel move the spent fuel underwater from the reactor to the pool. Over time, as the spent fuel is stored in the pool, it becomes cooler as the radioactivity decays away. After several years (> 5 years), decay heat and radioactivity decreases under specified limits so that spent fuel may be reprocessed or interim storaged.

As was written, shielding of the spent fuel pool is ensured by:

  • the design of the spent fuel pool (water and concrete shielding),
  • the design of surrounding working area
  • requirements on water level in the pool,

Spent fuel pools are fitted with stainless steel and aluminum racks that hold the fuel assemblies and are lined with stainless steel to prevent leaking. There are no drains that would allow the water level to drop or the pool to become empty. The plants have a variety of extra water sources and equipment to replenish water that evaporates over time, or in case there is a leak. Plant personnel are also trained and prepared to quickly respond to a problem. The water serves two purposes: it cools the fuel and shields workers at the plant from radioactivity. Although water is neither high density nor high Z material, it is commonly used as gamma shields. Water provides a radiation shielding of fuel assemblies in a spent fuel pool during storage or during transports from and into the reactor core. Although water is a low-density material and low Z material, it is commonly used in nuclear power plants, because these disadvantages can be compensated with increased thickness.

The minimum water level in the fuel storage pool meets primarily the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel. But it also provides with the coolant required for cooling spent fuel.

References:
Nuclear and Reactor Physics:
  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
  3. D. L. Hetrick, Dynamics of Nuclear Reactors, American Nuclear Society, 1993, ISBN: 0-894-48453-2.
  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See also:

Spent Fuel Pool

We hope, this article, Safety of Spent Fuel Pool, helps you. If so, give us a like in the sidebar. Main purpose of this website is to help the public to learn some interesting and important information about physics and reactor physics.

What is Spent Fuel Pool – Definition

Spent fuel pool (SFP) is storage pool for spent nuclear fuel from nuclear reactors. Spent fuel pool may be located inside the containment building or inside the fuel building. Reactor Physics
Spent fuel pool
Spent fuel pool. Source: wikipedia.org License: Public Domain

Spent fuel pool (SFP) is storage pool for spent nuclear fuel from nuclear reactors. Spent fuel pool may be located inside the containment building or inside the fuel building (outside the containment building). When located outside the containment building, the two areas are connected by a fuel transfer system which carries the fuel through a normally closed opening in the reactor containment. In this case spent fuel is removed from the reactor vessel by a manipulator crane and placed in the fuel transfer system. In the spent fuel pool, the fuel is removed from the transfer system and placed into storage racks. After a suitable decay period, the fuel can be removed from storage and loaded into a shipping cask for removal from the site. Spent fuel pools are typically 12m or more deep, with the bottom equipped with storage racks designed to hold fuel assemblies removed from the reactor. A reactor’s pool is specially designed for the reactor in which the fuel was used and situated at the reactor site.

Spent fuel pools are fitted with stainless steel and aluminum racks that hold the fuel assemblies and are lined with stainless steel to prevent leaking. There are no drains that would allow the water level to drop or the pool to become empty. The plants have a variety of extra water sources and equipment to replenish water that evaporates over time, or in case there is a leak. Plant personnel are also trained and prepared to quickly respond to a problem. The water serves two purposes: it cools the fuel and shields workers at the plant from radioactivity. Although water is neither high density nor high Z material, it is commonly used as gamma shields. Water provides a radiation shielding of fuel assemblies in a spent fuel pool during storage or during transports from and into the reactor core. Although water is a low-density material and low Z material, it is commonly used in nuclear power plants, because these disadvantages can be compensated with increased thickness.

To conserve space, in all plants open storage racks were replaced with so called high-density racks that incorporate (boron-10) or other neutron-absorbing material to ensure subcriticality. Using such racks, fuel assemblies can be stored in about one half the volume required for storage in standard racks.

Spent Nuclear Fuel
Spent Fuel - Fuel Assembly
Typical fuel assembly

Spent nuclear fuel, also called the used nuclear fuel, is a nuclear fuel that has been irradiated in a nuclear reactor (usually at a nuclear power plant or an experimental reactor) and that must be replaced by a fresh fuel due to its insufficient reactivity. Spent nuclear fuel is characterized by fuel burnup which is a measure of how much energy is extracted from a nuclear fuel and a measure of fuel depletion. Due to fuel depletion and fission fragments buildup, spent nuclear fuel is no longer useful in sustaining a nuclear reaction in an ordinary thermal reactor and it must be replaced by fresh fuel. Depending on its point along the nuclear fuel cycle, it may have considerably different isotopic constituents.

It must be noted, irradiated fuel is due to presence of high amount of radioactive fission fragments and transuranic elements very hot and very radioactive. Reactor operators have to manage the heat and radioactivity that remains in the “spent fuel” after it’s taken out of the reactor.  In nuclear power plants, spent nuclear fuel is usually stored underwater in the spent fuel pool on the plant.  Plant personnel move the spent fuel underwater from the reactor to the pool. Over time, as the spent fuel is stored in the pool, it becomes cooler as the radioactivity decays away. After several years (> 5 years), decay heat decreases under specified limits so that spent fuel may be reprocessed or interim storaged.

At first glance, it is difficult to recognize a fresh fuel from an used fuel. From mechanical point of view, the used fuel (irradiated) is identical as the fresh fuel. In most PWRs, used fuel assemblies stand between four and five metres high, are about 20 cm across and weighs about half a tonne. A PWR fuel assembly comprises a bottom nozzle into which rods are fixed through the lattice and to finish the whole assembly it is ended by a top nozzle. There are spacing grids between these nozzles. These grids ensure an exact guiding of the fuel rods. The bottom and top nozzles are heavily constructed as they provide much of the mechanical support for the fuel assembly structure. Western PWRs use a square lattice arrangement and assemblies are characterized by the number of rods they contain, typically, 17×17 in current designs. In contrast to the fresh fuel, which are simply shiny, the oxide layer forming on the surface of used fuel assemblies during the four-year fuel cycle makes them dark. Moreover, Cherenkov radiation is typical only for spent nuclear fuel. The glow is visible also after the chain reaction stops (in the reactor). The cherenkov radiation can characterize the remaining radioactivity of spent nuclear fuel, therefore it can be used for measuring of fuel burnup.

Safety of Spent Fuel Pool

Safety of spent fuel pools stands on various criteria. These criteria may be grouped according to following aspects:

  • Subcriticality. Fulfillment of this criterion is based on:
    • the design of the spent fuel pool,
    • requirements on boric acid diluted in water,
    • limiting of stored fuel (e.g. fuel enrichment, assembly burnup)
  • Cooling. Fulfillment of this criterion is based on:
    • the design of the spent fuel pool (e.g. no drains below the top of the stored fuel elements),
    • requirements on water level in the pool,
    • requirements on active cooling elements (heat exchangers and heat sink).
  • Radiation Shielding. Fulfillment of this criterion is based on:
    • the design of the spent fuel pool (water and concrete shielding),
    • the design of surrounding working area
    • requirements on water level in the pool,
  • Integrity. Fulfillment of this criterion is based on:
    • the design of the spent fuel pool,
    • the design of surrounding working area,
    • ensuring periodic inspections

Subcriticality of Spent Fuel Pool

As was written, subcriticality of the spent fuel pool is ensured by:

  • the design of the spent fuel pool,
  • requirements on boric acid diluted in water,
  • limiting of stored fuel (e.g. fuel enrichment, assembly burnup)

Today, spent fuel is usually stored in so called high-density racks or in the maximum-density rack (MDR).  Using such racks, fuel assemblies can be stored in about one half the volume required for storage in standard racks. Higher storage densities have been achieved without the risk of a nuclear chain reaction by adding neutron absorbing materials (typically boron) in storage racks and baskets, and dissolved in the water itself. These racks incorporate (boron-10) or other neutron-absorbing material to ensure subcriticality. Boron-10 is generally present in the chemical form of boron carbide (B4C) within a metal matrix (e.g., Boral and Metamic (trademark of Metamic, LLC)) or a polymer matrix (e.g., Boraflex (trademark of BISCO), Carborundum, and Tetrabor), although borated stainless steel incorporates the boron-10 atoms into the alloy composition.

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. In Region 2, maximum reactivity fuel is not allowed to load, while it can be loaded in the racks of Region 1 of the pool.

Most conservative approach requires that the multiplication factor, assuming flooding with pure water and infinite geometry, does not exceed 0.95 with a full loading of the maximum anticipated enrichment. To satisfy this design criterion, the assumptions in the criticality evaluation are as below:

  • The fuel assemblies have the maximum approved initial enrichment with the highest reactivity in fuel’s lifetime, and without the control rods (burnable poisons may be taken into account).
  • In the flooding condition, all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 (5% subcriticality) be evaluated in the absence of soluble boron.
  • The array of the fuel assemblies can be taken as infinite geometry, or with reflective boundary condition.
  • The effect of structure material and the fixed neutron absorber can be considered.
  • Unless the double contingency principle is taken, the presence of the boron in the moderation should not be considered. This principle shows at least two independent, unlikely and concurrent incidents have to happen to lead a criticality accident.

Boron Credit

The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. We mean that boric acid is dissolved in the coolant. Boric acid (molecular formula: H3BO3), is a white powder that is soluble in water. According to the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 (5% subcriticality) be evaluated in the absence of soluble boron. Hence, the design of the spent fuel pool is based on the use of unborated water. Unless the double contingency principle is taken, the presence of the boron (boron credit) in the moderation should not be considered. Nuclear plant owners are facing increasing fuel assembly enrichments, spent-fuel assembly burnup limitations, spent fuel pool storage cell restrictions, and problems with fixed neutron absorber degradation, all of which are challenging their traditional criticality analyses. The credit for soluble boron (boron credit or partial boron credit – PBC) in the spent fuel pool criticality analysis offers a solution to these concerns.

For example, according to NUREG-0800 (9.1.1-4):

“For PWR pools where partial credit for soluble boron is taken, both of the following criteria must be met:

  1. When the spent fuel storage racks are loaded with fuel of the maximum permissible reactivity and are flooded with full-density unborated water, the maximum K(eff) must be less than 1.0 for all normal and credible abnormal conditions. The K(eff) must include allowance for all relevant uncertainties and tolerances.
  2. When the spent fuel storage racks are loaded with fuel of the maximum permissible reactivity and are flooded with full-density water borated to a minimum concentration (CB,min, measured in parts per million of boron), the maximum K(eff) must be no greater than 0.95 for all normal conditions. Plant technical specifications must incorporate the CB,min. The K(eff) must include allowance for all relevant uncertainties and tolerances.”

Double Contingency Principle – Double Contingency Approach

“The double contingency approach requires a demonstration that unintended criticality cannot occur unless at least two unlikely, independent, concurrent changes in the conditions originally specified as essential to criticality safety have occurred.”

Source: Nuclear Safety Technical Assessment Guide. NS-TAST-GD-041 Revision 5. ONR, 2016.

The double contingency principle discussed in ANSI/ANS-8.1 allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the movement of fuel within spent fuel pool, and accidental misloading of a fuel assembly in the Region 2 of the spent fuel pool. This could potentially increase the criticality of Region 2. To mitigate these postulated criticality related accidents, boron is dissolved in the pool water.

Burnup Credit

In criticality licensing of spent fuel pool, taking credit for the decrease in fuel reactivity due to fuel burnup is known as burnup credit. Burnup credit is similar to boron credit. The concept of burnup credit is taking credit for the reduction in reactivity due to irradiation of nuclear fuel when the criticality safety analysis is carried out for the spent fuel. In spent fuel storage, it has generally been required that for criticality analyses that it be assumed that fuel si at its peak reactivity, which is generally the fresh fuel. But in the spent fuel pool, most of fuel assemblies have higher burnup with lower reactivity.

The reduction of reactivity is a combinative effect of:

  • the net reduction of fissile nuclides,
  • the production of neutron-absorbing nuclides (non-fissile actinides and fission products)

In the high-density rack design, the spent fuel storage pool may divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. There are two kinds of storage racks:

  • The storage racks applied to the spent fuel assemblies for which the burnup values must be equal or more than the burnup value credited in the criticality safety calculation. (Region 2)
  • The storage racks applied to the spent fuel assemblies assuming non-irradiated, with the maximum reactivity (Region 1).

It should be clear that the burnup credit is not attempt to reduce the safety margins in criticality safety. It is just to reduce the analysis conservatism, in another word, reduce the uncertainties in safety margins by a more accurate safety analysis. The fresh fuel assumption can be very conservative and result in a significant reduction in capacity for a given storage or cask volume. The issue is complicated by ensuring that adequate administrative controls and measurement systems exist to prevent higher reactivity fuel from being placed in storage racks. Regulators also have concerns about verifying the accuracy of being allowed burnup credit.

For example, abnormal conditions should include consideration of fuel assemblies loaded into storage racks not approved for their storage (e.g., fuel not meeting minimum burnup requirements stored in burnup credit storage racks).

Special reference: Introduction of Burn-up Credit in Nuclear Criticality Safety Analysis. Guoshun You, Chunming Zhang, Xinyi Pan. Nuclear and Radiation Safety Center of MEP.

Cooling of Spent Fuel Pool

Decay HeatIt must be noted, irradiated fuel is due to presence of high amount of radioactive fission fragments and transuranic elements very hot and very radioactive. Reactor operators have to manage the heat and radioactivity that remains in the “spent fuel” after it’s taken out of the reactor.  In nuclear power plants, spent nuclear fuel is usually stored underwater in the spent fuel pool on the plant. Plant personnel move the spent fuel underwater from the reactor to the pool. Over time, as the spent fuel is stored in the pool, it becomes cooler as the radioactivity decays away. After several years (> 5 years), decay heat decreases under specified limits so that spent fuel may be reprocessed or interim storaged.

However, even at these low levels (about 0.25% after ten days), the amount of heat generated requires the continued removal of heat for an appreciable time after shutdown. Decay heat is a long-term consideration and impacts spent fuel handling, reprocessing, waste management, and reactor safety.

The design of the reactor and the spent fuel pool must allow for the removal of this decay heat from the core by some means. If adequate heat removal is not available, decay heat will increase the temperatures in the fuel to the point that fuel melting and fuel assembly damage will occur as in the case of Fukushima. The degree of concern with decay heat will vary according to reactor type and design.

As was written, cooling of the spent fuel pool is ensured by:

  • the design of the spent fuel pool (e.g. no drains below the top of the stored fuel elements),
  • requirements on water level in the pool,
  • requirements on active cooling elements (heat exchangers and heat sink).

The minimum water level in the fuel storage pool meets primarily the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel. But it also provides with the coolant required for cooling spent fuel.

Cooling of the spent fuel pool is usually ensured by the spent fuel pool cooling system, which is designed to remove residual decay heat generated by spent fuel stored in the spent fuel pool. The system also maintains the purity of the spent fuel cooling water and the refueling water. The system is designed to provide with cooling of stored spent fuel and it may become necessary to cool all fuel assemblies from totally unloaded reactor. The design incorporates redundant active components (usually 3×100%) and it is one of safety systems. One of principal safety functions of the ultimate heat sink (UHS) is dissipation of maximum expected decay heat from the spent fuel pool to ensure the pool temperature remains within the design bounds for the structure.

Shielding of Spent Fuel in Spent Fuel Pool

It must be noted, irradiated fuel is due to presence of high amount of radioactive fission fragments and transuranic elements very hot and very radioactive. Reactor operators have to manage the heat and radioactivity that remains in the “spent fuel” after it’s taken out of the reactor.  In nuclear power plants, spent nuclear fuel is usually stored underwater in the spent fuel pool on the plant. Plant personnel move the spent fuel underwater from the reactor to the pool. Over time, as the spent fuel is stored in the pool, it becomes cooler as the radioactivity decays away. After several years (> 5 years), decay heat and radioactivity decreases under specified limits so that spent fuel may be reprocessed or interim storaged.

As was written, shielding of the spent fuel pool is ensured by:

  • the design of the spent fuel pool (water and concrete shielding),
  • the design of surrounding working area
  • requirements on water level in the pool,

Spent fuel pools are fitted with stainless steel and aluminum racks that hold the fuel assemblies and are lined with stainless steel to prevent leaking. There are no drains that would allow the water level to drop or the pool to become empty. The plants have a variety of extra water sources and equipment to replenish water that evaporates over time, or in case there is a leak. Plant personnel are also trained and prepared to quickly respond to a problem. The water serves two purposes: it cools the fuel and shields workers at the plant from radioactivity. Although water is neither high density nor high Z material, it is commonly used as gamma shields. Water provides a radiation shielding of fuel assemblies in a spent fuel pool during storage or during transports from and into the reactor core. Although water is a low-density material and low Z material, it is commonly used in nuclear power plants, because these disadvantages can be compensated with increased thickness.

The minimum water level in the fuel storage pool meets primarily the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel. But it also provides with the coolant required for cooling spent fuel.

Spent Fuel – Cherenkov Radiation

Cherenkov Radiation in the reactor core.
Cherenkov Radiation in the reactor core.

The cherenkov radiation is electromagnetic radiation emitted when a charged particle (such as an electron) moves through a dielectric medium faster than the phase velocity of light in that medium.

Cherenkov radiation can be used to detect high-energy charged particles (especially beta particles). In nuclear reactors or in a spent nuclear fuel pool, beta particles (high-energy electrons) are released as the fission fragments decay. The glow is visible also after the chain reaction stops (in the reactor). The cherenkov radiation can characterize the remaining radioactivity of spent nuclear fuel, therefore it can be used for measuring of fuel burnup.

References:
Nuclear and Reactor Physics:
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Advanced Reactor Physics:

  1. K. O. Ott, W. A. Bezella, Introductory Nuclear Reactor Statics, American Nuclear Society, Revised edition (1989), 1989, ISBN: 0-894-48033-2.
  2. K. O. Ott, R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, 1985, ISBN: 0-894-48029-4.
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  4. E. E. Lewis, W. F. Miller, Computational Methods of Neutron Transport, American Nuclear Society, 1993, ISBN: 0-894-48452-4.

See also:

Nuclear Power Plant

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