Nuclear Reactor

In general, a nuclear reactor is a device used to initiate and control a self-sustained nuclear chain reaction. Since nuclear reactors are used in nuclear power plants, research facilities or nuclear-propelled ships, their output is thermal energy or they can be used as a source of neutron radiation.

From the physics point of view, the main differences among reactor types arise from differences in their neutron energy spectra. In fact, the basic classification of nuclear reactors is based upon the average energy of the neutrons which cause the bulk of the fissions in the reactor core. From this point of view nuclear reactors are divided into two categories:

• Thermal Reactors. Almost all of the current reactors which have been built to date use thermal neutrons to sustain the chain reaction. These reactors contain neutron moderator that slows neutrons from fission until their kinetic energy is more or less in thermal equilibrium with the atoms (E < 1 eV) in the system.
• Fast Neutron Reactors. Fast reactors contains no neutron moderator and use less-moderating primary coolants, because they use fast neutrons (E > 1 keV), to cause fission in their fuel.

The main conventional types of nuclear reactors are:

Most common nuclear reactors are light water reactors (LWR), where light water is used as a moderator. LWR’s are divided into two categories:

• Pressurized water reactors (PWR) – are characterized by high pressure primary circuit (to keep the water in liquid state)
• Boiling water reactors (BWR) – are characterized by controlled boiling in the primary circuit

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).  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%.

See also: Types of Nuclear Reactors

The key components common to most PWR-type nuclear reactors are:

• 

  Nuclear reactor core. The reactor core is a bounded region, where a neutron multiplication occurs and where chain reaction take place. The reactor core contains especially the nuclear fuel (fuel assemblies), the moderator and the control rods.

• Reactor pressure vessel. The reactor pressure vessel is the pressure vessel containing the reactor core and other key reactor internals. It is a cylindrical vessel with a hemispherical bottom head and a flanged and gasketed upper head. The bottom head is welded to the cylindrical shell while the top head is bolted to the cylindrical shell via the flanges. The top head is removable to allow for the refueling of the reactor during planned outages.
• Neutron reflector. The reactor core is surrounded by a neutron reflector or reactor core baffle. The reflector reduces the non-uniformity of the power distribution in the peripheral fuel assemblies, reduces neutron leakage and reduces a coolant flow bypass of the core.
• Lower support structure.
• Surveillance capsule assembly. The capsules contain reactor vessel steel specimens obtained during vessel fabrication and are withdrawn periodically from the reactor vessel.
• Core support barrel. The core barrel belongs to the lower core support structure, because it houses a reactor core.
• Upper guide structure assembly. 

In general, the reactor thermal power and the outlet temperature of the coolant from the reactor core are controlled by manipulating several factors which affect the core’s reactivity. In PWRs, these factors are especially:

• position of control rods,
• concentration of boric acid in the RCS
• core inlet temperature

These three paramenters determines reactivity of the reactor system. A control rod is removed from or inserted into the reactor core in order to increase or decrease the reactivity of the reactor (increase or decrease the neutron flux).

See also: Reactor Control

Reactor Criticality

The basic classification of states of a reactor is according to the multiplication factor as eigenvalue which is a measure of the change in the fission neutron population from one neutron generation to the subsequent generation.

• 

  k_eff < 1. If the multiplication factor for a multiplying system is less than 1.0, then the number of neutrons is decreasing in time (with the mean generation time) and the chain reaction will never be self-sustaining. This condition is known as the subcritical state.

• k_eff = 1. If the multiplication factor for a multiplying system is equal to 1.0, then there is no change in neutron population in time and the chain reaction will be self-sustaining. This condition is known as the critical state.

• k_eff > 1. If the multiplication factor for a multiplying system is greater than 1.0, then the multiplying system produces more neutrons than are needed to be self-sustaining. The number of neutrons is exponentially increasing in time (with the mean generation time). This condition is known as the supercritical state.

See also: Reactor Criticality

Example – How to Change Power of Nuclear Reactor

During any power increase the temperature, pressure, or void fraction change and the reactivity of the core changes accordingly. It is difficult to change any operating parameter and not affect every other property of the core. Since it is difficult to separate all these effects (moderator, fuel, void etc.) the power coefficient is defined. The power coefficient combines the Doppler, moderator temperature, and void coefficients. It is expressed as a change in reactivity per change in percent power, Δρ/Δ% power. The value of the power coefficient is always negative in core life but is more negative at the end of the cycle primarily due to the decrease in the moderator temperature coefficient.

Let assume that the reactor is critical at 75% of rated power and that the plant operator wants to increase power to 100% of rated power. The reactor operator must first bring the reactor supercritical by insertion of a positive reactivity (e.g. by control rod withdrawal or borondilution). As the thermal power increases, moderator temperature and fuel temperature increase, causing a negative reactivity effect (from the power coefficient) and the reactor returns to the critical condition. In order to keep the power to be increasing, positive reactivity must be continuously inserted (via control rods or chemical shim). After each reactivity insertion, the reactor power stabilize itself proportionately to the reactivity inserted. The total amount of feedback reactivity that must be offset by control rod withdrawal or boron dilution during the power increase (from ~1% – 100%) is known as the power defect.

Let assume:

• the power coefficient:                 Δρ/Δ% = -20pcm/% of rated power
• differential worth of control rods:    Δρ/Δstep = 10pcm/step
• worth of boric acid:                                      -11pcm/ppm
• desired trend of power decrease:              1% per minute

75% → ↑ 20 steps or ↓ 18 ppm of boric acid within 10 minutes → 85% → next ↑ 20 steps or ↓ 18 ppm within 10 minutes → 95% → final ↑ 10 steps or ↓ 9 ppm within 5 minutes → 100%

Consumption of a 3000MWth (~1000MWe) reactor (12-months fuel cycle)

It is an illustrative example, following data do not correspond to any reactor design.

• Typical reactor may contain about 165 tonnes of fuel (including structural material)
• Typical reactor may contain about 100 tonnes of enriched uranium (i.e. about 113 tonnes of uranium dioxide).
• This fuel is loaded within, for example, 157 fuel assemblies composed of over 45,000 fuel rods.
• A common fuel assembly contain energy for approximately 4 years of operation at full power.
• Therefore about one quarter of the core is yearly removed to spent fuel pool (i.e. about 40 fuel assemblies), while the remainder is rearranged to a location in the core better suited to its remaining level of enrichment (see Power Distribution).
• The removed fuel (spent nuclear fuel) still contains about 96% of reusable material (it must be removed due to decreasing k_inf of an assembly).

• Annual natural uranium consumption of this reactor is about 250 tonnes of natural uranium (to produce of about 25 tonnes of enriched uranium).

• Annual enriched uranium consumption of this reactor is about 25 tonnes of enriched uranium.

• Annual fissile material consumption of this reactor is about 1 005 kg.

• Annual matter consumption of this reactor is about 1.051 kg.

• But it corresponds to about 3 200 000 tons of coal burned in coal-fired power plant per year.

See also: Fuel Consumption

The main problems or rather challenges, that must be taken into account when designing reactors, are:

• Pressure and temperature stresses with associated limits
  • Pressure and Temperature (P/T) Limits
  • Heatup and cooldown rates
  • Low-temperature Overpressure Protection Limits
  • Pressurized Thermal Shock
• Radiation Damage to Reactor Materials
  • Ductile-brittle Transition Temperature
  • Upper-shelf Toughness
• Corrosion

See also: Material Problems of Nuclear Reactors

Special Reference: Reactor Pressure Vessel Status Report, U.S. NRC. NUREG-1511. Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission, Washington, 1994.

Did you know?

The world’s first nuclear reactor operated about two billion years ago. The natural nuclear reactor formed at Oklo in Gabon, Africa, when a uranium-rich mineral deposit became flooded with groundwater that acted as a neutron moderator, and a nuclear chain reaction started.  These fission reactions were sustained for hundreds of thousands of years, until a chain reaction could no longer be supported. This was confirmed by existence of isotopes of the fission-product gas xenon and by different ratio of U235/U238 (enrichment of natural uranium).

The existence of this phenomenon was discovered in 1972 at Oklo in Gabon, Africa.

See previous:

Reactor Types

See above:

Nuclear Power Plant

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Reactor Core