Boiling water reactor

The Boiling Water Reactor (BWR) is a type of nuclear reactor used for the generation of electrical power. It is the second most common type of electricity-generating nuclear reactor after the pressurized water reactor (PWR). The BWR was developed by the Idaho National Laboratory and General Electric in the mid-1950s. In the present, General Electric specializes in the design and construction of this type of reactor.

Overview

The BWR uses ordinary water (light water) as a coolant and neutron moderator. Heat is produced by nuclear fission in the reactor core, and this causes the cooling water to boil, producing steam. The steam is directly used to drive a turbine, after which is cooled in a condenser and converted back to liquid water. This water is then returned to the reactor core, completing the loop. The cooling water is maintained at about 75 times atmospheric pressure so that it boils in the core at about 285°C. In comparison, there is no significant boiling allowed in a PWR because of the high pressure maintained in its primary loop (about 158 times atmospheric pressure).

Description of Major Components and Systems

Feedwater

Steam exiting from the turbine flows into condensers located underneath the low pressure turbines where the steam is cooled and returned to the liquid state (condensate). The condensate is then pumped through feedwater heaters that raise its temperature using extraction steam from various turbine stages. Feedwater from the feedwater heaters enters the reactor pressure vessel (RPV) through nozzles high on the vessel, well above the top of the nuclear fuel assemblies (these nuclear fuel assemblies constitute the "core") but below the water level.

The feedwater enters into the downcomer region and combines with water exiting the water separators. The feedwater subcools the saturated water from the steam separators. This water now flows down the downcomer region, which is separated from the core by a tall shroud. The water then goes through either jet pumps or internal recirculation pumps that provide additional pumping power (hydraulic head). The water now makes a 180 degree turn and moves up through the lower core plate into the nuclear core where the fuel elements heat the water. Water exiting the fuel channels at the top guide is about 12 to 15% saturated steam (by mass), typical core flow may be 100E6 lb/hr with 14.5E6 lb/hr steam flow. However, core-average void fraction is a significantly higher fraction (~40%). These sort of values may be found in each plant's publicly available Technical Specifications, Final Safety Analysis Report, or Core Operating Limits Report.

The heating from the core creates a thermal head that assists the recirculation pumps in recirculating the water inside of the RPV. A BWR can be designed with no recirculation pumps and rely entirely on the thermal head to recirculate the water inside of the RPV. The forced recirculation head from the recirculation pumps is very useful in controlling power, however. The thermal power level is easily varied by simply increasing or decreasing the forced recirculation flow through the recirculation pumps.

The two phase fluid (water and steam) above the core enters the riser area, which is the upper region contained inside of the shroud. The height of this region may be increased to increase the thermal natural recirculation pumping head. At the top of the riser area is the water separator. By swirling the two phase flow in cyclone separators, the steam is separated and rises upwards towards the steam dryer while the water remains behind and flows horizontally out into the downcomer region. In the downcomer region, it combines with the feedwater flow and the cycle repeats.

The saturated steam that rises above the separator is dried by a chevron dryer structure. The steam then exits the RPV through four main steam lines and goes to the turbine.

Control systems

Reactor power is controlled via two methods: by inserting or withdrawing control rods and by changing the water flow through the reactor core.

Positioning (withdrawing or inserting) control rods is the normal method for controlling power when starting up a BWR. As control rods are withdrawn, neutron absorption decreases in the control material and increases in the fuel, so reactor power increases. As control rods are inserted, neutron absorption increases in the control material and decreases in the fuel, so reactor power decreases. Some early BWRs and the proposed ESBWR (Economic Simplified BWR) designs use only natural circulation with control rod positioning to control power from zero to 100% because they do not have reactor recirculation systems. Fine reactivity adjustment would be accomplished by modulating feedwater temperatute/flow???

Changing (increasing or decreasing) the flow of water through the core is the normal and convenient method for controlling power. When operating on the so-called "100% rod line," power may be varied from approximately 30% to 100% of rated power by changing the reactor recirculation system flow by varying the speed of the recirculation pumps. As flow of water through the core is increased, steam bubbles ("voids") are more quickly removed from the core, the amount of liquid water in the core increases, neutron moderation increases, more neutrons are slowed down to be absorbed by the fuel, and reactor power increases. As flow of water through the core is decreased, steam voids remain longer in the core, the amount of liquid water in the core decreases, neutron moderation decreases, fewer neutrons are slowed down to be absorbed by the fuel, and reactor power decreases.

Steam Turbines

Steam produced in the reactor core passes through steam separators and dryer plates above the core and then directly to the turbine, which is part of the reactor circuit. Because the water around the core of a reactor is always contaminated with traces of radionuclides, the turbine must be shielded during normal operation, and radiological protection must be provided during maintenance. The increased cost related to operation and maintenance of a BWR tends to balance the savings due to the simpler design and greater thermal efficiency of a BWR when compared with a PWR. Most of the radioactivity in the water is very short-lived (mostly N-16, with a 7-second half-life), so the turbine hall can be entered soon after the reactor is shut down.

Size

A modern BWR fuel assembly comprises 74 to 100 fuel rods, and there are up to approximately 800 assemblies in a reactor core, holding up to approximately 140 tonnes of uranium. The number of fuel assemblies in a specific reactor is based on considerations of desired reactor power output, reactor core size and reactor power density.

Safety Systems

Like the pressurized water reactor, the BWR reactor core continues to produce heat from radioactive decay after the fission reactions have stopped, making nuclear meltdown possible in the event that all safety systems have failed and the core does not receive coolant. Also like the pressurized water reactor, a boiling-water reactor has a negative void coefficient, that is, the thermal output decreases as the proportion of steam to liquid water increases inside the reactor. However, unlike a pressurized water reactor which contains no steam in the reactor core, a sudden increase in BWR steam pressure (caused, for example, by a blockage of steam flow from the reactor) will result in a sudden decrease in the proportion of steam to liquid water inside the reactor. The increased ratio of water to steam will lead to increased neutron moderation, which in turn will cause an increase in the power output of the reactor. Because of this effect in BWRs, operating components and safety systems are designed to ensure that no credible, postulated failure can cause a pressure and power increase that exceeds the safety systems' capability to quickly shutdown the reactor before damage to the fuel or to components containing the reactor coolant can occur.

In the event of an emergency that disables all of the safety systems, each reactor is surrounded by a containment building designed to seal off the reactor from the environment.

Reactor Protection System (SCRAM)

Depending on the power level of the reactor (i.e. ascending or at power) there are safety-related contingencies that may arise that necessitate a rapid emergency shutdown of the reactor, or, in Western nuclear parlance, a "SCRAM".^{(See note .)} The SCRAM is a manually-triggered or automatically-triggered rapid insertion of all control rods into the reactor, which will take the reactor to decay heat power levels within tens of seconds. Since ~ 0.6% of neutrons are emitted from fission products ("delayed" neutrons), which are born seconds/minutes after fission, all fission can not be terminated instantaneously. Manual SCRAMs may be initiated by the reactor operators; while automatic SCRAMs are initiated upon:

1. Low reactor water level indicative of:
   1. loss of coolant accident (i.e. LOCA)
   2. loss of proper feedwater (LOFW)
   3. etc.
2. High drywell (primary containment) pressure
   1. indicative of loss of coolant accident
3. Main Steam Isolation Valve Closure (MSIV)
   1. indicative of main steam line break
4. Turbine stop valve or turbine control valve closure
   1. if turbine protection systems wish to cease admission of steam the Reactor SCRAM is in anticipation of a pressure transient that would increase reactivity (collapse boiling voids)
   2. generator load rejection will also cause closure of turbine valves and SCRAM reactor
5. Loss of Offsite Power (LOOP)
   1. during normal operation, the reactor protection system (RPS) is powered by offsite power
      1. loss of offsite power would open all relays in the RPS would open causing all SCRAM signals to come in redundantly
      2. would also cause MSIV to close since RPS is fail safe; plant assumes a main steam break is coincident with loss of offsite power

Evolution of the BWR

Early Concepts

The BWR concept was developed slightly later than the PWR concept. Development of the BWR started in the mid-1950s, and was a collaboration between GE and several US national laboratories.

The "test" (as opposed to "production") BWRs of the 1950s through early/mid-1960s only partially used directly-generated (primary) nuclear boiler system steam to feed the turbine and incorporated heat exchangers for the generation of secondary steam to drive seperate parts of the the turbines. The literature does not indicate why this was the case, but it was eliminated on production models of the BWR.

First Series of Production BWRs (BWR1-BWR6)

The first generation of production boiling water reactors saw the incremental development of the unique and distinctive features of the BWR, such as the torus, used to quench steam in the event of a transient requiring the quenching of steam, as well as the drywell, the elimination of the heat exchanger, the steam dryer, and the distinctive general layout of the reactor building, as well as the standardization of reactor control and safety systems. The first series of production BWRs evolved through 6 iterative design phases, each termed BWR-1 through BWR-6. (BWR-4s and BWR-6s are the most common types in service today.) The vast majority of BWRs in service throughout the world belong to one of these design phases.

The Advanced Boiling Water Reactor (ABWR)

A newer design of BWR is known as the Advanced Boiling Water Reactor (ABWR). The ABWR was developed in the late 1980s and early 1990s, and has been further improved to the present day. The ABWR incorporates advanced technologies in the design, including computer control, plant automation, control rod removal, motion, and insertion, in-core pumping, and nuclear safety to deliver improvements over the original series of production BWRs, with a high power output (1350 MWe/reactor), and a significantly lowered probability of core damage. Most significantly, the ABWR was a completely standardized design, that could be made for mass-production (insofar as any extraordinarily complex system can be mass-produced.)

The ABWR was approved by the U.S. Nuclear Regulatory Commission for production as a standardized design in the early 1990s. Subsequently, numerous ABWRs were built in Japan; these have reportedly performed with distinction, both safely and economically, in Japanese service. One development spurred by the success of the ABWR in Japan is that GE's nuclear energy division merged with Hitachi Corporation's nuclear energy division, forming GE Hitachi, who is now the major worldwide developer of the BWR design.

The Simplified Boiling Water Reactor (SBWR)

GE also developed a different concept for a new BWR at the same time as the ABWR, known as the Simplified Boiling Water Reactor (SBWR). Designed at a time when nuclear power was under enhanced scrutiny due to a series of safety contingencies in operating plants as well as regulatory pressure, this smaller (600 MWe/reactor) was notable for its incorporation - for the first time ever in a light water reactor - of "passive safety" design principles. The concept of passive safety means that the reactor, rather than requiring the intervention of active systems, such as emergency injection pumps, to keep the reactor within safety margins, was instead designed to return to a safe state if a safety-related contingency developed solely through operation of natural forces.

For example, if the reactor got too hot, it would melt a seal that would release liquid "nuclear poisons", or materials that greatly hamper a chain reaction by absorbing neutrons, into the reactor core. The tank containing the "nuclear poisons" would be located above the reactor, and the poisons, once the seal was melted, would flow into the core through force of gravity, and bring the reaction to a near-complete stop. Another example was the Isolation Condenser system, which relied on the principle of hot water/steam rising to bring hot coolant into large heat exchangers located above the reactor in very deep tanks of water, thus accomplishing residual heat removal. Yet another example was the omission of recirculation pumps within the core; these pumps were used in other BWR designs to keep cooling water moving; they were expensive, hard to reach to repair, and could occasionally fail; so as to improve reliability, the ABWR incorporated no less than 10 of these recirculation pumps, so that even if several failed, a sufficient number would remain serviceable so that an unscheduled shutdown would not be necessary, and the pumps could be repaired during the next refuelling outage. Instead, the designers of the Simplified Boiling Water Reactor used thermal analysis to design the reactor core such that natural circulation (cold water falls, hot water rises) would bring water to the center of the core to be boiled.

The ultimate result of the passive safety features of the SBWR would be a reactor that, assuming everything worked as planned, would not require human intervention in the event of a major safety contingency for at least 48 hours following the safety contingency; thence, it would only require periodic refilling of cooling water tanks located completely outside of the reactor, isolated from the cooling system, and designed to remove reactor waste heat through evaporation. The Simplified Boiling Water Reactor was not submitted to or approved by the NRC, or even built; still, the concept remained intriguing to GE's designers, and served as the basis of future developments.

The Economic Simplified Boiling Water Reactor (ESBWR)

During a period beginning in the late 1990s, GE engineers proposed to combine the features of the Advanced Boiling Water Reactor design with the distinctive safety features of the Simplified Boiling Water Reactor design, along with scaling up the resulting design to a larger size of 1550 MWe (4500 MWth). This design has been submitted to the U.S. Nuclear Regulatory Commission for approval, and the subsequent Final Design Review is near completion.

Reportedly, this design has a best-in-class core damage probability of 3 x 10^-8 core damage events per reactor-year. (In essence, this means that for each ESBWR built, there is a 50% chance that a partial or complete meltdown will occur if the reactor is operated for 333 1/3 million years. Earlier designs of the BWR (the BWR/4) had core damage probabilities as high as a 50% chance of of partial or complete meltdown for every hundred thousand years of operation or 1 x 10^-5 core damage events per reactor-year.)

Several reactors of this design are now on order within the United States.

Advantages and Disadvantages

Advantages

• The reactor vessel and associated components operate at a substantially lower pressure (about 75 times atmospheric pressure) compared to a PWR (about 158 times atmospheric pressure).
• Pressure vessel is subject to significantly less irradiation compared to a PWR, and so does not become as brittle with age.
• Operates at a lower nuclear fuel temperature.
• Fewer components due to no steam generators and no pressurizer vessel. (Older BWRs have external recirculation loops, but even this piping is eliminated in modern BWRs, such as the ABWR.)
• Lower risk (probability) of a rupture causing loss of coolant compared to a PWR, and lower risk of a severe accident should such a rupture occur. This is due to fewer pipes, fewer large diameter pipes, fewer welds and no steam generator tubes.
• Measuring the water level in the pressure vessel is the same for both normal and emergency operations, which results in easy and intuitive assessment of emergency conditions.
• Can operate at lower core power density levels using natural circulation without forced flow.
• A BWR may be designed to operate using only natural circulation so that recirculation pumps are eliminated entirely. (The new ESBWR design uses natural circulation.)
• BWRs do not use boric acid to control fission burn-up, leading to less possibility of corrosion within the reactor vessel and piping. (Corrosion from boric acid must be carefully monitored in PWRs; it has been demonstrated that dangerous reactor vessel head corrosion can occur if the reactor vessel head is not properly maintained. See Davis-Besse. Since BWRs do not utilize boric acid, these contingencies are eliminated.)

Disadvantages

• Complex calculations for managing consumption of nuclear fuel during operation due to "two phase (water and steam) fluid flow" in the upper part of the core. This requires more instrumentation in the reactor core. The innovation of computers, however, makes this less of an issue.
• Much larger pressure vessel than for a PWR of similar power, with correspondingly higher cost. (However, the overall cost is reduced because a modern BWR has no main steam generators and associated piping.)
• Contamination of the turbine by short-lived activation products. This means that shielding and access control around the steam turbine are required during normal operations due to the radiation levels arising from the steam entering directly from the reactor core.
• Control rods are inserted from below for current BWR designs. There are two available hydraulic power sources that can drive the control rods into the core for a BWR under emergency conditions. There is a dedicated high pressure hydraulic accumulator and also the pressure inside of the reactor pressure vessel available to each control rod. Either the dedicated accumulator (one per rod) or reactor pressure is capable of fully inserting each rod. Most other reactor types use top entry control rods that are held up in the withdrawn position by electromagnets, causing them to fall into the reactor by gravity if power is lost.

Technical and Background Information

Start-Up ("Going Critical")

Prior to the introduction of the Fine Motion Control Rod Drive with the ABWR and ESBWR, control rod motion could not be controlled in a boiling water reactor with smooth motion, but instead, control rods moved through a series of notched positions with fixed intervals between these positions, though control rods could be controlled individually or in banks?. So as to assure a smooth start-up, GE developed a set of rules in the 1970s called BPWS (Banked Position Withdrawal Sequence) that help minimize notch worths and aided in starting the reactor using asymmetric control rod withdrawal patterns. (For instance, rather than withdrawing all control rods from 50% to 49%, which was impossible without fine-grained control, 20% of the control rods would be lowered from 50% to 45%, giving an effective level of 49% for the reactor.)

Thermal Margins

Three calculated/measured quantities are tracked while operating a BWR:

• Maximum Fraction Limiting Critical Power Ratio, or MFLCPR;
• Fraction Limiting Linear Heat Generation Rate, or FLLHGR;
• Average Planar Linear Heat Generation Rate, or APLHGR;

All three of these quantities must be kept less than 1.0 during normal operation; administrative controls are in place to assure some margin of error and margin of safety to these licensed limits. Typical computer simulations divide the reactor core into 24-25 axial planes; relevant quantities (margins, burnup, power, void history) are tracked for each "node" in the reactor core (764 fuel assemblies x 25 nodes/assembly = 19100 nodal calculations/quantity).

Maximum Fraction Limiting Critical Power Ratio (MFLCPR)

Specifically, MFLCPR represents how close the leading fuel bundle is to "dry-out" or "departure from nucleate boiling." Transition boiling is the unstable transient region where nucleate boiling tends toward film boiling. A water drop dancing on a hot frying pan is an example of film boiling. During film boiling a volume of insulating vapor separates the heated surface from the cooling fluid; this causes the temperature of the heated surface to increase drastically to once again reach equilibrium heat transfer with the cooling fluid. In other words, steam semi-insulates the heated surface and surface temperature rises to allow heat to get to the cooling fluid (through convection and radiative heat transfer).

MFLCPR is monitored with an empirical correlation that is formulated by vendors of BWR fuel (GE, Westinghouse, AREVA-NP). The vendors have test rigs where they simulate nuclear heat with resistive heating and determine experimentally what conditions of coolant flow, fuel assembly power, and reactor pressure will be in/out of the transition boiling region for a particular fuel design. In essence, the vendors make a model of the fuel assembly but power it with resistive heaters. These mock fuel assemblies are put into a test stand where data points are taken at specific powers, flows, pressures. It is obvious that nuclear fuel could be damaged by film boiling; this would cause the fuel cladding to overheat and fail. Experimental data is conservatively applied to BWR fuel to ensure that the transition to film boiling does not occur during normal or transient operation. Typical SLMCPR/MCPRSL (Safety Limit MCPR) licensing limit for a BWR core is substantiated by a calculation that proves that 99.4% of fuel rods in a BWR core will not enter the transition to film boiling in the event of the worst possible plant transient/SCRAM anticipated to occur. Since the BWR is boiling water, and steam does not transfer heat as well as liquid water, MFCLPR typically occurs at the top of a fuel assembly, where steam volume is the highest.

Fraction Limiting Linear Heat Generation Rate (FLLHGR)

FLLHGR (FDLRX, MFLPD) is a limit on fuel rod power in the reactor core. For new fuel, this limit is typically around 13 Kw/foot of fuel rod. This limit ensures that the centerline temperature of the fuel pellets in the rods will not exceed the melting point of the fuel material (uranium/gadolinium oxides) in the event of the worst possible plant transient/scram anticipated to occur. To illustrate the response of LHGR in transient imagine the rapid closure of the valves that admit steam to the turbines at full power. This causes the immediate cessation of steam flow and an immediate rise in BWR pressure. This rise in pressure effectively subcools the reactor coolant instantaneously; the voids (vapor) collapse into solid water. When the voids collapse in the reactor, the fission reaction is encouraged (more thermal neutrons); power increases drastically (120%) until it is terminated by the automatic insertion of the control rods. So, when the reactor is isolated from the turbine rapidly, pressure in the vessel rises rapidly, which collapses the water vapor, which causes a power excursion which is terminated by the Reactor Protection System. If a fuel pin was operating at 13.0 Kw/foot prior to the transient, the void collapse would cause its power to rise. The FLLHGR limit is in place to ensure that the highest powered fuel rod will not melt if its power was rapidly increased following a pressurization transient. Abiding by the LHGR limit precludes melting of fuel in a pressurization transient.

Average Planar Linear Heat Generation Rate (APLHGR)

APLHGR, being an average of LHGR, is a margin associated with fuel melting during a LOCA (Loss of Coolant Accident - a catastrophic loss of coolant pressure within the reactor, considered the "primary design basis threat" in probabilistic risk assessment and nuclear safety). BWR designs incorporate failsafe protection systems (the ESBWR has completely passively safe emergency systems, while the older models of BWR incorporate active systems) that will ensure the integrity of the reactor fuel in the event of a massive pipe rupture and rapid de-pressurization of the vessel, which would uncover the fuel. These protection systems have capacities that they can handle and it is required that the heat stored in the fuel assemblies at any one time does not overwhelm the protection systems, such as the Emergency Feedwater Injection/Emergency Core Cooling System, which inject massive quantities of water into the reactor vessel, flooding the core, and cooling it. APLHGR is monitored to ensure that the reactor is not operated at an average power level that would defeat the primary containment systems. When a refueled core is licensed to operate, the fuel vendor/licensee simulate transients with computer models. Their approach is to simulate worst case transients in the reactor's most vulnerable states.

List of BWRs

U.S. Commercial Boiling Water Reactor Nuclear Power Plants

• Big Rock Point, Michigan (decommissioned)
• BONUS, Puerto Rico (decommissioned) [1]
• Browns Ferry Nuclear Plant (Reactors 1, 2, and 3)
• Brunswick Nuclear Generating Station, North Carolina
• Clinton Nuclear Generating Station, Illinois
• Columbia Nuclear Generating Station, Washington (aka WNP-2, Hanford-2, WPPSS-2)
• Cooper Nuclear Station, Nebraska
• Dresden Nuclear Power Plant, Illinois
• Duane Arnold Energy Center, Iowa
• Elk River Station, Minnesota (decommissioned)
• Enrico Fermi Unit 2, Michigan
• Fitzpatrick Nuclear Generating Station, New York
• Grand Gulf Nuclear Generating Station, Mississippi
• Hatch (Edwin I. Hatch) Nuclear Generating Station, Georgia
• Hope Creek Nuclear Generating Station, New Jersey
• Humboldt Bay, California (decommissioned)
• La Crosse Boiling Water Reactor, Wisconsin (decommissioned)
• LaSalle County Nuclear Generating Station, Illinois
• Limerick Nuclear Power Plant, Pennsylvania
• Millstone Nuclear Power Plant (Reactor 1 only), Connecticut (decommissioned)
• Monticello Nuclear Generating Plant, Minnesota
• Nine Mile Point Nuclear Generating Station, New York
• Oyster Creek Nuclear Generating Station, New Jersey
• Peach Bottom Nuclear Generating Station, Pennsylvania
• Perry Nuclear Generating Station, Ohio
• Pilgrim Nuclear Generating Station, Massachusetts
• Quad Cities Nuclear Generating Station, Illinois
• River Bend Nuclear Generating Station, Louisiana
• Shoreham Nuclear Generating Station, New York (decommissioned)
• Susquehanna Steam Electric Station, Pennsylvania
• Vallecitos Nuclear Center, California (idle)
• Vermont Yankee, Vermont

Other commercial BWRs

Commercial BWRs outside the USA include:

• Finland:
  • Olkiluoto units 1 & 2
• Germany:
  • Brunsbüttel
  • Gundremmingen A (permanently shut down)
  • Gundremmingen B & C
  • Isar unit 1
  • Krümmel
  • Lingen (permanently shut down)
  • Philippsburg unit 1.
  • Würgassen KKW (permanently shut down)
• India:
  • Tarapur units 1 & 2
• Japan:
  • Tokai JPDR (decommissioned)
  • Fukushima Daiichi units 1-6
  • Fukushima Daini units 1-4
  • Hamaoka units 1-4

• • Shimane units 1 & 2
  • Tokai unit 2
  • Tsuruga unit 1
• Netherlands
  • Dodewaard (permanently shut down)
• Mexico:
  • Veracruz, Laguna Verde, units 1 & 2
• Spain:
  • Cofrentes (1 unit)
  • Santa María de Garoña (1 unit)
• Sweden:
  • Barsebäck units 1 & 2 [2] (permanently shut down)
  • Forsmark units 1-3
  • Oskarshamn units 1-3
  • Ringhals unit 1 (units 2-4 are PWRs)
• Switzerland:
  • Leibstadt (1 unit)
  • Mühleberg (1 unit)

Experimental and other BWRs

Experimental and other non-commercial BWRs include:

• SL-1 (permanently shut down following accident in 1961)

Next-generation designs

• Advanced Boiling Water Reactor (ABWR)
• Economic Simplified Boiling Water Reactor (ESBWR)

See also

• List of nuclear reactors
• SL-1 Accident and Lessons
• Containment building
• Nuclear Power 2010 Program

External links

• Boiling Water Reactor, BWR Simulator Program
• Boiling Water Reactors, US Nuclear Regulatory Commission
• Advanced BWR General Description (Table of Contents, with active links to text)
• Technical details and features of Advanced BWRs
• Text Book Chapter on "Nuclear Power Reactors" describing various reactor types
• The Nuclear Tourist website