- Pressurized water reactor – PWR
- Boiling water reactor – BWR
- Manufacturing of the Reactor Vessel – Youtube
- Fuel Consumption – Summary
- Components of a nuclear reactor
- Reactor Pressure Vessel
- Material Problems and Challenges of Nuclear Reactors
- Core Barrel
- Neutron Reflector
- How to Change Power of Reactor
- Reactor Criticality
- Criticality of a Power Reactor – Power Defect
- Power Distribution in Conventional Reactor Cores
- Did you know?
- Bang Goes The Theory from Youtube
- at nuclear power plants for electricity generation
- at nuclear research facilities as a neutron source
- as a propulsion of nuclear propelled ships.
- 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 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). 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%.
Boiling water reactor – BWR
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.
See also: Boiling water reactor
Manufacturing of the Reactor Vessel – YoutubeDescription of VVER-1000 reactor.
- Control Element Drive Mechanism
- Reactor vessel head assembly
- Reactor pressure vessel
- Coolant inlet – outlet nozzles
- Downcomer for coolant
- Neutron reflector
- Fuel assemblies
Fuel Consumption – Summary
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 kinf 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
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.
There is one inlet (or cold leg) nozzle and one outlet (or hot leg) nozzle for each reactor coolant system loop. The reactor coolant enters the reactor vessel at the inlet nozzle and exits the reactor at the upper internals region, where it is routed out the outlet nozzle into the hot leg of primary circuit and goes on to the steam generator. The primary circuit of typical PWR is divided into 4 independent loops (piping diameter ~ 800mm), each loop comprises a steam generator and one main coolant pump, but can differ according to certain reactor design. Therefore numerous inlet and outlet nozzles, as well as control rod drive tubes (in case of BWRs) and instrumentation and safety injection nozzles penetrate the cylindrical shell. This number of inlet and outlet nozzles is a function of the number of loops.
The body of the reactor vessel is constructed of a high-quality low-alloy carbon steel, and all surfaces that come into contact with reactor coolant are clad with a minimum of about 3 to 10 mm of austenitic stainless steel in order to minimize corrosion.
The reactor pressure vessels are the highest priority key components in nuclear power plants. The reactor pressure vessel houses the reactor core and because of its function it has direct safety significance. During the operation of a nuclear power plant, the material of the reactor pressure vessel is exposed to neutron radiation (especially to fast neutrons), which results in localized embrittlement of the steel and welds in the area of the reactor core. In order to minimize such material degradation radial neutron reflectors are installed around the reactor core. There are two basic types of neutron reflectors, the core baffle and the heavy reflector. Due to higher atomic number density heavy reflectors reduce neutron leakage (especially of fast neutrons) from the core more efficiently than core baffle. Since the reactor pressure vessel is considered irreplaceable, these ageing effects of the RPV have the potential to be life-limiting conditions for a nuclear power plant.
Material Problems and Challenges of Nuclear Reactors
The main problems or rather challenges, that must be taken into account when designing reactors, are:
- Pressure and temperature stresses with associated limits
- Radiation Damage to Reactor Materials
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.
In general, reactor internals are divided into three structural units:
- the lower core support structure
- the upper core support structure
- the in-core instrumentation
The core barrel belongs to the lower core support structure, because it houses a reactor core. Other lower core support structures (lower core plate, core baffle or heavy reflector) are attached to the core barrel, which transmits the weight of the core to the reactor vessel. The barrel is a long, cylindrical, one-piece welded structure. Like most components of the internals, the core barrel is made of low carbon, chromium-nickel stainless steel, because it is situated in a corrosive environment (primary coolant comprises boric acid), and the material should not get oxidized.
The lower internals and also the core barrel remain in place during refueling, but may be removed for reactor pressure vessel in-service inspections.
It is well known that each 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. The neutron reflector is a non-multiplying medium, whereas the reactor core is a multiplying medium.
The neutron reflector scatters back (or reflects) into the core many neutrons that would otherwise escape (i.e. reduces the neutron leakage). By reducing neutron leakage, the reflector increases keff and reduces the amount of fuel necessary to maintain the reactor critical for a long period. In LWRs the neutron reflector is installed for following purposes:
- The neutron flux distribution is “flattened“, i.e., the ratio of the average flux to the maximum flux is increased. Therefore reflectors reduce the non-uniformity of the power distribution.
- Because of the higher flux at the edge of the core, there is much better utilization in the peripheral fuel assemblies. This fuel, in the outer regions of the core, now contributes much more to the total power production.
- The neutron reflector scatters back (or reflects) into the core many neutrons that would otherwise escape. The neutrons reflected back into the core are available for chain reaction. This means that the minimum critical size of the reactor is reduced. Alternatively, if the core size is maintained, the reflector makes additional reactivity available for higher fuel burnup. The decrease in the critical size of core required is known as the reflector savings.
- Neutron reflectors reduce neutron leakage i.e. to reduce the neutron fluence on a reactor pressure vessel.
- Neutron reflectors reduce a coolant flow bypass of a core.
- Neutron reflectors serve as a thermal and radiation shield of a reactor core.
See also: Neutron Reflector
How to Change Power of Reactor
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.
keff < 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.
- keff = 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.
- keff > 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.
Criticality of a Power Reactor – Power Defect
For power reactors at power conditions the reactor can behave differently (in comparison to zero power reactor) as a result of the presence of reactivity feedbacks. Power reactors are initially started up from hot standby mode (subcritical state at 0% of rated power) to power operation mode (100% of rated power) by withdrawing control rods and by boron dilution from the primary coolant. During the reactor startup and up to about 1% of rated power, the reactor kinetics is exponential as in zero power reactor. This is due to the fact all temperature reactivity effects are minimal.
On the other hand, during further power increase from about 1% up to 100% of rated power, the temperature reactivity effects play very important role. As the neutron population increases, the fuel and the moderator increase its temperature, which results in decrease in reactivity of the reactor (almost all reactors are designed to have the temperature coefficients negative).
The negative reactivity coefficient acts against the initial positive reactivity insertion and this positive reactivity is offset by negative reactivity from temperature feedbacks. 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 on the power level 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 is known as the power defect. The power defects for PWRs, graphite-moderated reactors, and sodium-cooled fast reactors are:
- about 2500pcm for PWRs,
- about 800pcm for graphite-moderated reactors
- about 500pcm for sodium-cooled fast reactors
The power defects slightly depend on the fuel burnup, because they are determined by the power coefficient which depends on the fuel burnup. 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.
It is logical, as power defects act against power increase, they act also against power decrease. When reactor power is decreased quickly, as in the case of reactor trip, power defect causes a positive reactivity insertion, and the initial rod insertion must be sufficient to make the reactor safe subcritical. It is obvious, if the power defect for PWRs is about 2500pcm (about 6 βeff), the control rods must weigh more than 2500pcm to achieve the subcritical condition. To ensure the safe subcritical condition, the control rods must weigh more than 2500pcm plus value of SDM (SHUTDOWN MARGIN). The total weigh of control rods is design specific, but, for example, it may reach about 6000pcm. To ensure that the control rods can safe shut down the reactor, they must be maintained above a minimum rod height (rods insertion limits).
Power Distribution in Conventional Reactor Cores
It should be noted the flux shape derived from the diffusion theory is only a theoretical case in a uniform homogeneous cylindrical reactor at low power levels (at “zero power criticality”). We have implicitly assumed that the core consisting of thousands of fuel and control elements, coolant, and structure can be represented by some effective homogeneous mixture. This is a very strong assumption, because it does not take into account the heterogeneity of a core.
See also: Heterogeneous Core
In commercial reactor cores the flux distribution is significantly influenced by:
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.