Fast Neutron Reactor

Fast spectrum vs. Thermal spectrum
Neutron cross-section for fission.
Fast spectrum vs. Thermal spectrum
Source: www.world-nuclear.org
Nuclear Power -> Nuclear Power Plant -> Types of Reactors -> Fast Neutron Reactor

A fast neutron reactor is a nuclear reactor in which the fission chain reaction is sustained by fast neutrons. That means the neutron moderator (slowing down) in such reactors is undesirable. This is a key advantage of fast reactors, because fast reactors have a significant excess of neutrons (due to low parasitic absorbtion), unlike PWRs (or LWRs).

Sodium-cooled Fast Reactor (SFR).
Sodium-cooled Fast Reactor (SFR).
Source: wikipedia.org

On the other hand such reactors must compensate for the missing reactivity from neutron moderator efect. They use fuel with higher enrichment when compared to that required for a thermal reactor. Fast reactors require enrichments about 10%, or more. Most fast reactors use a hexagonal lattice cells (as VVER reactors) in order to reach smaller volume ratios of coolant to fuel. Generally, fast reactors have to utilize much more compact nuclear cores than thermal reactors (PWRs or BWRs) in order to reach required core reactivity. This implies the fast reactor cores achieve higher power densities. As a consequence, they cannot use water as coolant, because of its moderating properties and insufficient thermal properties. The solution given this problem is to use another coolant as liquid sodium or lead.

Lead-cooled Fast Reactor (LFR)
Lead-cooled Fast Reactor (LFR).
Source: wikipedia.org

Fast reactor fuel may be metal or a ceramic, encapsulated in metal cladding, unlike the PWR’s zirconium cladding. Liquid metals are the most widely used coolant because they have excellent heat transfer properties and can be employed in lowpressure systems. Sodium-cooled fast reactors (SFRs) are the most common designs. Because sodium reacts violently with water, however, SFRs require the placement of an intermediate heat exchanger between the reactor core and the steam generator. This hi-tech technology requires a lot of experience, therefore only few countries have developed their own fast reactor design (e.g. Russia, USA, France, Japan, ). Especially Russians continue in fast reactor developement program with their BN reactors.

Breeder reactor

Neutrons can breed fuel
Free neutrons can “breed” more fuel from otherwise non-fissionable isotopes.
Source: hyperphysics.phy-astr.gsu.edu

A breeder reactor is essentially a particular configuration of a fast reactor (but not only FBR can be used as a breeder). Fast reactors generally have an excess of neutrons (due to low parasitic absorbtion), the neutrons given off by fission reactions can “breed” more fuel from otherwise non-fissionable isotopes or can be used for another purposes (e.g.transmutation of spent nuclear fuel). The most common breeding reaction is an absorbtion reaction on uranium-238, where a plutonium-239 from non-fissionable uranium-238 is produced. A key parameter of breeder reactors is a breeding ratio, although this ratio describes also thermal reactors fuel cycle.

The term “breeder” refers to the types of configurations which can be the breeding ratio higher than 1. That means such reactors produce more fissionable fuel than they consume (i.e. more fissionable Pu-239 is produced from non-fissionable uranium-238, than consumed initial U-235+Pu-239 fuel).

See also: Breeder Reactor

Advantages

  • FBRs have improved neutron economy
  • FBRs can recycle nuclear waste
  • FBRs can produce fuel for thermal reactors
  • FBRs liquid metals have superior heat transfer properties
  • FBRs do not use pressure vessel

Disadvantages

  • FBRs must use superior control system
  • FBRs can have positive reactivity feedback from void coefficient
  • Liquid metals require special technology and handling
  • Fast reactor technology can be more expensive

Generation IV reactors

In 2003 the Generation IV International Forum (GIF) representing ten countries announced the selection of six reactor technologies which they believe represent the future shape of nuclear energy. These were selected on the basis of being clean, safe and cost-effective means of meeting increased energy. Three of the six reactors are fast reactors and one can be built as a fast reactor, one is described as epithermal. Only two operate with slow neutrons like today’s plants.

 

Liquid Metal cooled Fast Reactors designs. Integral (pool) design vs. Loop design
Liquid Metal cooled Fast Reactors designs. Integral (pool) design vs. Loop design
Source: wikipedia.org

Breeder Reactor

Nuclear Power -> Nuclear Power Plant -> Types of Reactors -> Fast Neutron Reactor -> Breeder Reactor

Breeder Reactor

Fast reactor scheme.
Fast reactor scheme.
Source GE HITACHI

A breeder reactor is essentially a particular configuration of a fast reactor. Fast reactors generally have an excess of neutrons (due to low parasitic absorbtion), the neutrons given off by fission reactions can “breed” more fuel from otherwise non-fissionable isotopes or can be used for another purposes (e.g.transmutation of spent nuclear fuel). The most common breeding reaction is an absorbtion reaction on uranium-238, where a plutonium-239 from non-fissionable uranium-238 is produced. A key parameter of breeder reactors is a breeding ratio, although this ratio describes also thermal reactors fuel cycle.

The term “breeder” refers to the types of configurations which can be the breeding ratio higher than 1. That means such reactors produce more fissionable fuel than they consume (i.e. more fissionable Pu-239 is produced from non-fissionable uranium-238, than consumed initial U-235+Pu-239 fuel).

Production of fissile material in a reactor occurs by neutron irradiation of fertile material, particularly uranium-238 and thorium-232. These materials are breeded, either in the fuel or in a breeder blanket surrounding the core.

Pu-239 breeding. The uranium nucleus absorbs neutron, thus leads to Pu-239 breeding.
Pu-239 breeding. The uranium nucleus absorbs neutron, thus leads to Pu-239 breeding.
Source:

Types of breeder reactors

  • Fast breeder reactor (FBR). The superior neutron economy of a fast neutron reactor makes it possible to build a reactor that, after its initial fuel charge of plutonium, requires only natural (or even depleted) uranium feedstock as input to its fuel cycle. Russian BN-350 liquid-metal-cooled reactor was operated with a breeding ratio of over 1.2.
  • Thermal breeder reactor. The excellent neutron capture characteristics of fissile uranium-233 make it possible to build a moderated reactor that, after its initial fuel charge of enriched uranium, plutonium or MOX, requires only thorium as input to its fuel cycle. Thorium-232 produces uranium-233 after neutron capture and beta decay.

Breeding vs. Burnup

All commercial light water reactors breed fuel, but they have low breeding ratios. In recent years, the commercial power industry has been emphasizing high-burnup fuels (up to 60 – 70 GWd/tU), which are typically enriched to higher percentages of U-235 (up to 5%). As burnup increases, a higher percentage of the total power produced in a reactor is due to the fuel bred inside the reactor.

At a burnup of 30 GWd/tU (gigawatt-days per metric ton of uranium), about 30% of the total energy released comes from bred plutonium. At 40 GWd/tU, that percentage increases to about forty percent. This corresponds to a breeding ratio for these reactors of about 0.4 to 0.5. That means, about half of the fissile fuel in these reactors is bred there. This effect extends the cycle length for such fuels to sometimes nearly twice what it would be otherwise. MOX fuel has a smaller breeding effect than U-235 fuel and is thus more challenging and slightly less economic to use due to a quicker drop off in reactivity through cycle life.

AGR – Advanced Gas-cooled Reactor

Nuclear Power -> Nuclear Power Plant -> Types of Reactors -> Advanced Gas-cooled Reactor

An advanced gas-cooled reactor (AGR) is a British design of nuclear reactor. AGRs are using graphite as the neutron moderator and carbon dioxide as coolant. AGRs were developed from the Magnox type reactor.These are the second generation of British gas-cooled reactors. AGRs are operating at a higher gas temperature for improved thermal efficiency, thus requires stainless steel fuel cladding to withstand the higher temperature. Because the stainless steel fuel cladding has a higher neutron capture cross section than Magnox fuel (magnesium non-oxidising alloy), low enriched uranium fuel is needed.

The fuel is uranium oxide pellets, enriched to 2.5-3.5%, in stainless steel tubes. The carbon dioxide circulates through the core, reaching 650°C (for improved thermal efficiency) and then past steam generator tubes outside it, but still inside the concrete and steel pressure vessel (hence ‘integral’ design). Control rods penetrate the moderator and a secondary shutdown system involves injecting nitrogen to the coolant.

AGR - Advanced Gas-cooled Reactor
AGR – Advanced Gas-cooled Reactor
Source: www.hknuclear.com

 

CANDU – Heavy Water Reactor

Nuclear Power -> Nuclear Power Plant -> Types of Reactors -> CANDU - Heavy water reactor

The CANDU reactor design (or PHWR – Pressurized Heavy Water Reactor) has been developed since the 1950s in Canada, and more recently also in India. These reactors are heavy water cooled and moderated pressurized water reactors. Instead of using a single large reactor vessel as in a PWR or BWR, the nuclear core is contained in hundreds of pressure tubes. PHWRs generally use natural uranium (0.7% U-235) oxide as fuel, hence needs a more efficient moderator, in this case heavy water (D2O).

PHWR - CANDU
The PHWR design.
Source: www.cameco.com

The reactor core  is in a large tank called a calandria. There is a heavy water as the moderator in this tank. The calandria is penetrated by several hundred horizontal pressure tubes. These tubes  form channels for the fuel. The fuel is cooled by a flow of heavy water under high pressure in the primary cooling circuit, reaching 290°C. The moderator in the tank and the coolant in the channels are separated. The hot coolant that leaves the channels goes to a steam generator, which in turn heats a secondary loop of water to steam that can run turbines and generator (as in the PWR).

  • The PHWRs can be refueled while at full power, which makes them very efficient in their use of uranium (it allows for precise flux control in the core).
  • The PHWRs use a natural uranium as fuel (in the form of ceramic UO2).
  • The PHWRs produce more energy per kilogram of mined uranium than other designs, but also produces a much larger amount of used fuel per unit output.
  • The PHWRs produce a much larger amount of used fuel per unit output. (due to low fuel burnup)
  • Heavy water generally costs hundreds of dollars per kilogram, though this is a trade-off against reduced fuel costs.
  • Since the unenriched spent fuel generated less decay heat, the spent fuel can be stored much more compactly.
CANDU - fuel
CANDU – fuel
Source: www.cameco.com

 

 

PWR – Pressurized water reactor

Nuclear Power -> Nuclear Power Plant -> Types of Reactors -> Pressurized Water Reactor

Pressurized Water Reactor – PWR

Pressurized water reactors use a reactor pressure vessel (RPV) to contain the nuclear fuel, moderator, control rods and coolant. They are cooled and moderated by high-pressure liquid water (e.g. 16MPa). At this pressure water boils at approximately 350°C (662°F).  This high pressure is maintained by pressurizer. Inlet temperature of the water is about 290°C (554°F). The water (coolant) is heated in the reactor core to approximately 325°C (617°F) as the water flows through the core. As it can be seen, the reactor has approximately 25°C subcooled coolant (distance from the saturation).

The hot water that leaves the pressure vessel through hot leg nozzle and is looped through a steam generator, which in turn heats a secondary loop of water to steam that can run turbines and generator. Secondary water in the steam generator boils at pressure approximately 6-7 MPa, what equals to 260°C (500°F) saturated steam. Typical reactor nominal thermal power is about 3400MW, thus corresponds to the net electric output 1100MW. Therefor the typical efficiency of the Rankine cykle is about 33%.

The unused steam (45°C) is exhausted to the condenser, where it is condensed into water. The secondary side of the condenser extracts the waste heat (2000MW; 30°C) and the waste heat is released into environment. The resulting secondary water is pumped out of the condenser with a series of pumps, reheated, and pumped back to the steam generator.

Nuclear reactor - WWER 1200
Nuclear reactor and primary coolant system of WWER-1200.
Source: http://www.bellona.ru/

 

BWR – Boiling water reactor

Nuclear Power -> Nuclear Power Plant -> Types of Reactors -> Boiling Water Reactor

Boiling water reactor – BWR

ABWR boiling water reactor
A boiling water reactor (BWR) is cooled and moderated by water. It takes place at a lower pressure as in PWR, what allows the water to boil inside the pressure vessel producing the steam that runs the turbines.
Source: www.nuclearstreet.com

A boiling water reactor is cooled and moderated by water like a PWR, but at a lower pressure (7MPa), which allows the water to boil inside the pressure vessel producing the steam that runs the turbines. A BWR is like a PWR but with many differents.  The BWRs don’t have any steam generator. Unlike a PWR, there is no primary and secondary loop. The thermal efficiency of these reactors can be higher, and they can be simpler, and even potentially more stable and safe. But the disadvantage of this concept is that any fuel leak can make the water radioactive and that radioactivity can reach the turbine and the rest of the loop.

Electricity production

In a typical design concept of a BWRs, the following process occurs:

  • The reactor core inside the reactor vessel creates heat (e.g. 3900MW).
  • A steam-water mixture moves upward through the core and absorbs this heat (typical Tin=215°C and Tout=285°C).
  • The steam-water mixture leaves the top of the core and enters the two stages of moisture separation where water droplets are removed before the steam is allowed to enter the steamline.
  • The “dry” steam then exits the reactor through main steam lines and goes to the turbine.
  • The turbine drives the turbine generator, which produces electricity (e.g. 1300MWe).

The unused steam (45°C) is exhausted to the condenser, where it is condensed into water. The secondary side of the condenser extracts the waste heat (2600MW; 30°C) and the waste heat is released into environment. The resulting primary water is pumped out of the condenser with a series of pumps, reheated, and pumped back to the reactor vessel.

BWR simplified
Simlipied scheme of plant with boiling water reactor.
Source: www.cameco.com

 

 

 

 

Neutron Moderator

Neutron Moderators in Nuclear Reactors

The moderator, which is of importance in thermal reactors, is used to moderate, that is, to slow down, neutrons from fission to thermal energies. The probability that fission will occur depends on incident neutron energy. Physicists calculate with fission cross-section, which determines this probability.

Nuclei with low mass numbers are most effective for this purpose, so the moderator is always a low-mass-number material. In a fast reactor there is no moderator, only fuel and coolant. The moderation of neutrons is undesirable in fast reactors. Commonly used moderators include regular (light) water (roughly 75% of the world’s reactors), solid graphite (20% of reactors) and heavy water (5% of reactors). Beryllium and beryllium oxide (BeO) have been used occasionally, but they are very costly.

Why the moderator is needed?

The probability of the fission U-235 becomes very large at the thermal energies of slow neutrons. This fact implies increase of multiplication factor of the reactor (i.e. lower fuel enrichment is needed to sustain chain reaction)

Why fast reactors don’t need moderator?

Fast reactors use fast neutrons to split uranium or plutonium nuclei. They use higher fuel enrichment to sustain chain reaction. The moderation of neutrons is undesirable in fast reactors.

How does the neutron moderator work?
How does the neutron moderator work?
Source: http://hyperphysics.phy-astr.gsu.edu/
How does the neutron moderator work? Fuel pin
How does the neutron moderator work? Fuel pin
Source: http://hyperphysics.phy-astr.gsu.edu/

Elastic Scattering and Neutron Moderators

To be an effective moderator, the probability of elastic reaction between neutron and the nucleus must be high. In terms of cross-sections, the elastic scattering cross section of a moderator’s nucleus must be high. Therefore, a high elastic scattering cross-section is important, but does not describe comprehensively capabilities of moderators. In order to describe capabilities of a material to slow down neutrons, three new material variables must be defined:

Key properties of neutron moderators:
  • high cross-section for neutron scattering
  • high energy loss per collision
  • low cross-section for absorption
  • high melting and boiling point
  • high thermal conductivity
  • high specific heat capacity
  • low viscosity
  • low activity
  • low corrosive
  • cheap
 
Average Logarithmic Energy Decrement
During the scattering reaction, a fraction of the neutron’s kinetic energy is transferred to the nucleus. Using the laws of conservation of momentum and energy and the analogy of collisions of billiard balls for elastic scattering, it is possible to derive the following equation for the mass of target or moderator nucleus (M), energy of incident neutron (Ei) and the energy of scattered neutron (Es).

equation momentum energy

where A is the atomic mass number.

In case of the hydrogen (A = 1) as the target nucleus, the incident neutron can be completely stopped. But this works when the direction of the neutron is completely reversed (i.e. scattered at 180°). In reality, the direction of scattering ranges from 0 to 180 ° and the energy transferred also ranges from 0% to maximum. Therefore, the average energy of scattered neutron is taken as the average of energies with scattering angle 0 and 180°.

Moreover, it is useful to work with logarithmic quantities and therefore one defines the logarithmic energy decrement per collision (ξ) as a key material constant describing energy transfers during a neutron slowing down. ξ is not dependent on energy, only on A and is defined as follows:

logarithmic energy decrement - equation

For heavy target nuclei, ξ may be approximated by following formula:
the logarithmic energy decrement per collision

From these equations it is easy to determine the number of collisions required to slow down a neutron from, for example from 2 MeV to 1 eV.

Example:
Determine the number of collisions required for thermalization for the 2 MeV neutron in the carbon.
ξCARBON = 0.158
N(2MeV → 1eV) = ln 2⋅106/ξ =14.5/0.158 = 92

Table of average logarithmic energy decrement for some elements
Table of average logarithmic energy decrement for some elements.

For a mixture of isotopes:

the logarithmic energy decrement for mixtures

Macroscopic Slowing Down Power
We have defined the probability of elastic scattering reaction, we have defined the average energy loss during the reaction. The product of these variables (the logarithmic energy decrement and the macroscopic cross section for scattering in the material) is the macroscopic slowing down power (MSDP).

MSDP = ξ . Σs

The MSDP describes the ability of a given material to slow down neutrons and indicates how rapidly a neutron will slow down in the material, but it does not fully reflect the effectiveness of the material as a moderator. In fact, the material with high MSDP can slow down neutrons with high efficiency, but it can be a poor moderator because of its high probability of absorbing neutrons. It is typical, for example, for boron, which has a high slowing down power but is absolutely inappropriate as a moderator.

The most complete measure of the effectiveness of a moderator is the Moderating Ratio (MR), where:

MR  = ξ . Σs/Σa

Table of macroscopic slowing down power MSDP for some materials.
Table of macroscopic slowing down power MSDP for some materials.
Moderating Ratio
The moderating ratio or moderator quality is the most complete measure of the effectiveness of a moderator because it takes into account also the absorption effects. When absorption effects are high, most of the neutrons will be absorbed by moderator, leading to lower moderation or lower availability of thermal neutrons.

Therefore a higher ratio of MSDP to absorbtion cross sections ξ . Σs/Σa is desirable for effective moderation. This ratio is called the moderating ratio – MR and can be used as a criterion for comparison of different moderators.

Examples:

  • Light water has the highest ξ and σs among the moderators (resulting in the highest MSDP) shown in the table, but its moderating ratio is low due to its relatively higher absorption cross section.
  • On the other hand, heavy water has lower ξ and σs, but it has the highest moderating ratio owing to its lowest neutron absorption cross-section.
  • Graphite has much heavier nuclei than hydrogen in water, despite the fact graphite has much lower ξ and σs, it is better moderator than light water due to its lower absorption cross-section compared to that of light water.
Table of moderating ratios MR for some materials.
Table of moderating ratios for some materials.

See previous:

See above:

Neutron Reactions

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