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A nuclear reactorformerly known as reaktor 6 manual german free atomic pile so-called because the graphite moderator of the first reactor was placed into a tall pile [1]is a device used to reaktor 6 manual german free and control a fission nuclear chain reaction or nuclear fusion reactions.

Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion. Heat from nuclear fission is passed to a working fluid water or gasreaktor 6 manual german free in turn runs through steam turbines. These either drive a ship\’s propellers or turn electrical generators \’ shafts. Nuclear generated steam in principle can be used for industrial process heat or for district heating.

Some reactors are used to produce isotopes for medical and industrial use, or for production reaktor 6 manual german free weapons-grade plutonium. As of earlythe IAEA reports there are nuclear power reactors and nuclear research reactors in fred around the world.

Just as conventional thermal power stations generate electricity by harnessing the thermal energy released from burning fossil fuelsnuclear reactors convert the energy released by controlled nuclear fission into thermal energy for further conversion to mechanical or electrical forms. When a large fissile atomic nucleus such as uranium or plutonium absorbs a neutron, it may undergo nuclear fission.

The heavy nucleus splits into two or more lighter nuclei, the fission productsreleasing kinetic energygamma radiationand free neutrons. A portion of these neutrons may be absorbed by other fissile atoms and trigger further fission events, which release more neutrons, and so on. This is known as a nuclear chain reaction. To control such a nuclear chain reaction, control rods containing neutron poisons and neutron moderators reaktor 6 manual german free change the portion of neutrons that will go on to cause more fission.

A kilogram of uranium U converted via nuclear processes releases approximately three million times more energy than a kilogram of coal burned conventionally 7. The fission of one kilogram of uranium releases about 19 billion kilocaloriesso the energy released by 1 kg of uranium corresponds to that released by burning 2. Вот ссылка nuclear reactor coolant — usually water but sometimes a gas reaktor 6 manual german free a liquid metal like liquid sodium or lead or molten salt reaktor 6 manual german free is circulated past the reactor core to absorb the heat that it generates.

The heat is carried away from the reactor and is then used to generate steam. Most reactor systems employ a cooling system that is physically separated from the water that will be продолжить to produce pressurized steam for the turbineslike the pressurized water reactor.

However, in some reactors the water for the steam turbines is boiled directly by the reactor core ; for example the boiling water reactor. The rate of fission reactions within a reactor core can be adjusted by controlling the quantity of neutrons that are able to induce further fission events.

Nuclear reactors typically employ several methods of neutron control to adjust the reactor\’s power output. Reaktor 6 manual german free of these methods arise naturally from the physics of radioactive decay and are simply accounted for during the reactor\’s operation, while others are mechanisms engineered into the reactor design for a distinct purpose.

The fastest method for reajtor levels of fission-inducing neutrons in a reactor is reaktor 6 manual german free movement of the control rods. Control rods are made of neutron poisons reaktor 6 manual german free therefore absorb neutrons.

When a control rod is inserted deeper into the reactor, it absorbs more neutrons than the material it displaces — often the moderator. This action results in fewer neutrons available to cause fission and reduces the reactor\’s power output.

Conversely, extracting the control rod will result in an increase in the rate of fission events and reaktor 6 manual german free rree in power. The physics of radioactive decay also affects neutron populations in a reactor.

One yerman process is delayed neutron emission by a number of neutron-rich fission isotopes. These delayed neutrons account for about 0. The fission products which produce delayed neutrons have half-lives rraktor their decay by neutron emission that range reaktor 6 manual german free milliseconds to teaktor long as several minutes, and so considerable time is required to determine exactly when a reactor reaches the critical point.

Keeping the reactor in the zone of chain reactivity where delayed neutrons are necessary to achieve a critical mass state allows mechanical devices or human operators to control geramn chain reaction in \”real time\”; otherwise the time between achievement of germqn and nuclear meltdown as a result of an exponential power surge from the normal nuclear chain reaction, ger,an be too short to allow for intervention.

This last stage, where delayed neutrons are no longer required to maintain criticality, is known as the prompt critical point. There is a scale for describing criticality in numerical form, in which bare criticality is known as zero dollars and the prompt critical point is one dollarand other points in the process interpolated in cents. In some reactors, the coolant also acts as a neutron moderator. A moderator increases the power of the reactor by causing the fast neutrons that are released from fission to lose energy and become thermal neutrons.

Thermal neutrons are more likely than fast neutrons to cause fission. A higher temperature coolant would be less dense, and ftee a less effective moderator. In other reactors the coolant acts as a poison gegman absorbing neutrons in the same way that the control rods do. In these reactors power output can be increased by heating the coolant, which makes it a less dense poison. Nuclear reactors generally have automatic and reaitor systems to scram reaktor 6 manual german free reactor in an emergency shut down.

These systems insert large amounts of poison often boron in the form of boric acid into the reactor to shut the fission reaction down if unsafe conditions are detected or anticipated. Most types of reactors are sensitive to a process variously known as xenon poisoning, or the iodine pit. The common fission product Xenon produced in the fission process acts as a neutron poison that absorbs neutrons and therefore tends to shut the reactor down.

Xenon reaktor 6 manual german free can be controlled by keeping power levels high enough to destroy it by neutron absorption as fast as it is produced.

Fission also produces iodinewhich in turn decays with a half-life of 6. When the reactor is shut down, iodine continues to decay to xenon, making restarting the reactor more difficult for a day or two, as the xenon decays into cesium, which is not nearly as poisonous as xenon, with a half-life of 9.

This temporary state is the \”iodine pit. As the extra xenon is transmuted to xenon, which is much less a neutron poison, within a few hours the reactor experiences a \”xenon burnoff power transient\”. Control rods must be further inserted to replace the neutron absorption of the lost xenon Failure to properly follow such a procedure was a key step freee the Chernobyl disaster. Reactors used in nuclear marine propulsion especially nuclear submarines often cannot gerrman run reaktor 6 manual german free continuous power around the clock in reaktor 6 manual german free same way that land-based power reactors are normally run, and in addition often need to have a very long core life reaktor 6 manual german free refueling.

For this reason many designs use highly enriched uranium but incorporate burnable neutron poison in the fuel rods. The energy released in the fission process generates heat, some of which can be converted into usable energy. A common method of harnessing this thermal energy is to use it to boil water reaktor 6 manual german free produce pressurized steam which will then drive a steam turbine that turns an alternator and generates electricity.

The neutron was discovered in germam British physicist James Chadwick. He filed a patent for his idea of a simple reactor the following year while working at the Rraktor in London. Inspiration for a new type of reactor using uranium came from the discovery by Lise MeitnerFritz Strassmann and Otto Hahn in that bombardment of uranium with neutrons provided by an alpha-on-beryllium fusion reaction, a \” neutron howitzer \” produced a barium residue, which they reasoned was created by the fissioning of the uranium nuclei.

The U. The following year the U. Government received the Frisch—Peierls memorandum from the UK, which stated that the amount of uranium needed for a chain reaction was far lower than had previously been thought. Eventually, the first artificial nuclear reactor, Chicago Pile-1was constructed at the University of Chicagoby a team led by Italian physicist Enrico Fermiin late By this time, the program had been pressured for a year by U.

The reactor support structure was made of wood, which supported a pile hence the name of graphite blocks, embedded in which was natural uranium oxide \’pseudospheres\’ нажмите для продолжения \’briquettes\’. Soon after the Chicago Pile, the U. The reaktlr purpose for the largest reactors located at the Hanford Site in Washingtonwas berman mass production of plutonium for nuclear weapons.

Fermi and Szilard applied for a patent on reactors on 19 December Its issuance was delayed for 10 years because of wartime secrecy. Atomic Energy Commission produced 0. Besides the military uses of nuclear reactors, there were political reasons to pursue civilian use of atomic energy. This diplomacy led to the dissemination of reactor technology to U.

It produced around 5 MW electrical. It was built after the F-1 nuclear reactor which was the first reactor to go critical in Europe, and was also built by the Soviet Union. All commercial power reactors are based on nuclear fission. They generally use uranium and its product plutonium as nuclear fuelthough a thorium fuel cycle is also possible.

Fission reactors can be divided roughly into two classes, depending on the energy msnual the neutrons that sustain the fission chain reaction :. In principle, fusion power could be produced by nuclear fusion of elements such as the deuterium isotope of hydrogen. While an ongoing rich research topic since at least the s, no self-sustaining fusion germman for any purpose has ever been built. More than a dozen advanced reactor designs are in various stages of development.

Rolls-Royce aims to sell nuclear reactors for the production of synfuel for aircraft. Generation IV reactors are a set of theoretical nuclear reactor designs currently being researched. These designs are generally not expected to be available for commercial construction before Current reactors in operation around the world are generally considered second- or third-generation systems, with the first-generation systems having been retired some time ago. The primary goals being to improve nuclear safety, improve proliferation resistance, minimize waste and natural resource utilization, and to decrease the cost to build and run such plants.

Generation V reactors are designs which are theoretically possible, but which are not being actively considered or researched at present. Though some generation V reactors could potentially be built with current or near term technology, they trigger little interest for reasons of economics, practicality, or safety.

Controlled nuclear fusion could in principle be used in fusion power plants to produce power without the complexities of handling actinidesbut significant scientific and technical obstacles remain.

Several fusion reactors have been built, but reactors have never been able to release more energy than the amount of energy used in the process. Despite research having started in the s, no commercial fusion reactor is expected before The ITER project is currently leading the effort to harness fusion power. Thermal reactors generally depend on refined and enriched uranium.

Some nuclear reactors can operate with a mixture of plutonium and uranium see MOX. The process by which uranium ore is mined, processed, enriched, used, possibly reprocessed and disposed of is known as the nuclear fuel cycle. Enrichment involves increasing the percentage of U and is usually done by means of gaseous diffusion or gas centrifuge.

The enriched result is then converted into uranium dioxide больше информации, which is pressed reaktor 6 manual german free fired into pellet form. These pellets are stacked into tubes which are then sealed and called fuel rods. Many gdrman these fuel rods are used in each nuclear reactor. Theft risk of this fuel potentially used in the production of a nuclear weapon has led to campaigns advocating conversion of this type of reactor to low-enrichment uranium which poses less threat of proliferation.

Fissile U and non-fissile but fissionable and fertile U are both used in the fission process.

Soon after the Chicago Pile, the U. The primary purpose for the largest reactors located at the Hanford Site in Washington , was the mass production of plutonium for nuclear weapons. Fermi and Szilard applied for a patent on reactors on 19 December Its issuance was delayed for 10 years because of wartime secrecy. Atomic Energy Commission produced 0.

Besides the military uses of nuclear reactors, there were political reasons to pursue civilian use of atomic energy. This diplomacy led to the dissemination of reactor technology to U.

It produced around 5 MW electrical. It was built after the F-1 nuclear reactor which was the first reactor to go critical in Europe, and was also built by the Soviet Union.

All commercial power reactors are based on nuclear fission. They generally use uranium and its product plutonium as nuclear fuel , though a thorium fuel cycle is also possible.

Fission reactors can be divided roughly into two classes, depending on the energy of the neutrons that sustain the fission chain reaction :. In principle, fusion power could be produced by nuclear fusion of elements such as the deuterium isotope of hydrogen. While an ongoing rich research topic since at least the s, no self-sustaining fusion reactor for any purpose has ever been built.

More than a dozen advanced reactor designs are in various stages of development. Rolls-Royce aims to sell nuclear reactors for the production of synfuel for aircraft. Generation IV reactors are a set of theoretical nuclear reactor designs currently being researched. These designs are generally not expected to be available for commercial construction before Current reactors in operation around the world are generally considered second- or third-generation systems, with the first-generation systems having been retired some time ago.

The primary goals being to improve nuclear safety, improve proliferation resistance, minimize waste and natural resource utilization, and to decrease the cost to build and run such plants. Generation V reactors are designs which are theoretically possible, but which are not being actively considered or researched at present.

Though some generation V reactors could potentially be built with current or near term technology, they trigger little interest for reasons of economics, practicality, or safety.

Controlled nuclear fusion could in principle be used in fusion power plants to produce power without the complexities of handling actinides , but significant scientific and technical obstacles remain. Several fusion reactors have been built, but reactors have never been able to release more energy than the amount of energy used in the process. Despite research having started in the s, no commercial fusion reactor is expected before The ITER project is currently leading the effort to harness fusion power.

Thermal reactors generally depend on refined and enriched uranium. Some nuclear reactors can operate with a mixture of plutonium and uranium see MOX. The process by which uranium ore is mined, processed, enriched, used, possibly reprocessed and disposed of is known as the nuclear fuel cycle. Enrichment involves increasing the percentage of U and is usually done by means of gaseous diffusion or gas centrifuge.

The enriched result is then converted into uranium dioxide powder, which is pressed and fired into pellet form. These pellets are stacked into tubes which are then sealed and called fuel rods. Many of these fuel rods are used in each nuclear reactor.

Theft risk of this fuel potentially used in the production of a nuclear weapon has led to campaigns advocating conversion of this type of reactor to low-enrichment uranium which poses less threat of proliferation. Fissile U and non-fissile but fissionable and fertile U are both used in the fission process. U is fissionable by thermal i. A thermal neutron is one which is moving about the same speed as the atoms around it.

Since all atoms vibrate proportionally to their absolute temperature, a thermal neutron has the best opportunity to fission U when it is moving at this same vibrational speed. On the other hand, U is more likely to capture a neutron when the neutron is moving very fast. This U atom will soon decay into plutonium, which is another fuel. Pu is a viable fuel and must be accounted for even when a highly enriched uranium fuel is used.

Plutonium fissions will dominate the U fissions in some reactors, especially after the initial loading of U is spent.

Plutonium is fissionable with both fast and thermal neutrons, which make it ideal for either nuclear reactors or nuclear bombs. Most reactor designs in existence are thermal reactors and typically use water as a neutron moderator moderator means that it slows down the neutron to a thermal speed and as a coolant.

But in a fast breeder reactor , some other kind of coolant is used which will not moderate or slow the neutrons down much.

This enables fast neutrons to dominate, which can effectively be used to constantly replenish the fuel supply. By merely placing cheap unenriched uranium into such a core, the non-fissionable U will be turned into Pu, \”breeding\” fuel. In thorium fuel cycle thorium absorbs a neutron in either a fast or thermal reactor. The thorium beta decays to protactinium and then to uranium , which in turn is used as fuel.

Hence, like uranium , thorium is a fertile material. The amount of energy in the reservoir of nuclear fuel is frequently expressed in terms of \”full-power days,\” which is the number of hour periods days a reactor is scheduled for operation at full power output for the generation of heat energy. The number of full-power days in a reactor\’s operating cycle between refueling outage times is related to the amount of fissile uranium U contained in the fuel assemblies at the beginning of the cycle.

A higher percentage of U in the core at the beginning of a cycle will permit the reactor to be run for a greater number of full-power days. At the end of the operating cycle, the fuel in some of the assemblies is \”spent\”, having spent four to six years in the reactor producing power. This spent fuel is discharged and replaced with new fresh fuel assemblies. Plants typically operate on 18 month refueling cycles, or 24 month refueling cycles.

This means that one refueling, replacing only one-third of the fuel, can keep a nuclear reactor at full power for nearly two years. This nuclear waste is highly radioactive and its toxicity presents a danger for thousands of years.

The spent fuel pool is a large pool of water that provides cooling and shielding of the spent nuclear fuel. After loading into dry shielded casks, the casks are stored on-site in a specially guarded facility in impervious concrete bunkers. On-site fuel storage facilities are designed to withstand the impact of commercial airliners, with little to no damage to the spent fuel.

An average on-site fuel storage facility can hold 30 years of spent fuel in a space smaller than a football field. In a CANDU reactor, this also allows individual fuel elements to be situated within the reactor core that are best suited to the amount of U in the fuel element. The amount of energy extracted from nuclear fuel is called its burnup , which is expressed in terms of the heat energy produced per initial unit of fuel weight.

Burnup is commonly expressed as megawatt days thermal per metric ton of initial heavy metal. Nuclear safety covers the actions taken to prevent nuclear and radiation accidents and incidents or to limit their consequences. The nuclear power industry has improved the safety and performance of reactors, and has proposed new, safer but generally untested reactor designs but there is no guarantee that the reactors will be designed, built and operated correctly. Serious, though rare, nuclear and radiation accidents have occurred.

Nuclear reactors have been launched into Earth orbit at least 34 times. A number of incidents connected with the unmanned nuclear-reactor-powered Soviet RORSAT especially Kosmos radar satellite which resulted in nuclear fuel reentering the Earth\’s atmosphere from orbit and being dispersed in northern Canada January Almost two billion years ago a series of self-sustaining nuclear fission \”reactors\” self-assembled in the area now known as Oklo in Gabon , West Africa.

The conditions at that place and time allowed a natural nuclear fission to occur with circumstances that are similar to the conditions in a constructed nuclear reactor.

First discovered in by French physicist Francis Perrin , they are collectively known as the Oklo Fossil Reactors. Self-sustaining nuclear fission reactions took place in these reactors approximately 1. Such reactors can no longer form on Earth in its present geologic period. Radioactive decay of formerly more abundant uranium over the time span of hundreds of millions of years has reduced the proportion of this naturally occurring fissile isotope to below the amount required to sustain a chain reaction with only plain water as a moderator.

The natural nuclear reactors formed when a uranium-rich mineral deposit became inundated with groundwater that acted as a neutron moderator, and a strong chain reaction took place. The water moderator would boil away as the reaction increased, slowing it back down again and preventing a meltdown. The fission reaction was sustained for hundreds of thousands of years, cycling on the order of hours to a few days.

These natural reactors are extensively studied by scientists interested in geologic radioactive waste disposal. They offer a case study of how radioactive isotopes migrate through the Earth\’s crust. This is a significant area of controversy as opponents of geologic waste disposal fear that isotopes from stored waste could end up in water supplies or be carried into the environment.

Nuclear reactors produce tritium as part of normal operations, which is eventually released into the environment in trace quantities. As an isotope of hydrogen , tritium T frequently binds to oxygen and forms T 2 O.

This molecule is chemically identical to H 2 O and so is both colorless and odorless, however the additional neutrons in the hydrogen nuclei cause the tritium to undergo beta decay with a half-life of Despite being measurable, the tritium released by nuclear power plants is minimal.

The United States NRC estimates that a person drinking water for one year out of a well contaminated by what they would consider to be a significant tritiated water spill would receive a radiation dose of 0. The amounts of strontium released from nuclear power plants under normal operations is so low as to be undetectable above natural background radiation.

From Wikipedia, the free encyclopedia. Device used to initiate and control a nuclear chain reaction. This article is about constructed nuclear fission reactors. For nuclear fusion reactors, see Fusion power. For natural nuclear reactors, see Natural nuclear fission reactor. Main article: Nuclear reactor physics. Main article: Nuclear fission. PWR: Number of reactors by type end [23]. Net power capacity GWe by type end [23]. This section does not cite any sources.

Please help improve this section by adding citations to reliable sources. Unsourced material may be challenged and removed. June Learn how and when to remove this template message. Main article: Fusion power.

Main article: Nuclear fuel cycle. Main article: Nuclear safety. See also: Nuclear reactor safety system. See also: Lists of nuclear disasters and radioactive incidents. Main article: Natural nuclear fission reactor. Retrieved 2 August US Department of Energy.

Archived from the original PDF on 23 April Retrieved 24 September The Nuclear Tourist. Retrieved 25 September Archived from the original on 27 September Retrieved 18 March Cambridge University Press. ISBN Retrieved 17 March Archived from the original PDF on 11 December Krivit, Steven ed.

Hoboken, NJ: Wiley. American Nuclear Society Nuclear news. November Archived from the original PDF on 25 June Retrieved 18 June Retrieved 12 January BBC News. Retrieved 9 November At low power conditions, the feedwater controller acts as a simple PID control by watching reactor water level. At high power conditions, the controller is switched to a \”Three-Element\” control mode, where the controller looks at the current water level in the reactor, as well as the amount of water going in and the amount of steam leaving the reactor.

By using the water injection and steam flow rates, the feed water control system can rapidly anticipate water level deviations and respond to maintain water level within a few inches of set point. At this power level a single feedwater pump can maintain the core water level. If all feedwater is lost, the reactor will scram and the Emergency Core Cooling System is used to restore reactor water level. 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 due to neutron capture from the water, 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, with a 7-second half-life , so the turbine hall can be entered soon after the reactor is shut down. BWR steam turbines employ a high-pressure turbine designed to handle saturated steam, and multiple low-pressure turbines. The high-pressure turbine receives steam directly from the reactor.

The high-pressure turbine exhaust passes through a steam reheater which superheats the steam to over degrees F for the low-pressure turbines to use. The exhaust of the low-pressure turbines is sent to the main condenser. The steam reheaters take some of the turbine\’s steam and use it as a heating source to reheat what comes out of the high-pressure turbine exhaust.

While the reheaters take steam away from the turbine, the net result is that the reheaters improve the thermodynamic efficiency of the plant. A modern BWR fuel assembly comprises 74 to fuel rods , and there are up to approximately assemblies in a reactor core , holding up to approximately short tons of low-enriched 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.

A modern reactor has many safety systems that are designed with a defence in depth philosophy, which is a design philosophy that is integrated throughout construction and commissioning. A BWR is similar to a pressurized water reactor PWR in that the reactor will continue to produce heat even after the fission reactions have stopped, which could make a core damage incident possible.

This heat is produced by the radioactive decay of fission products and materials that have been activated by neutron absorption. BWRs contain multiple safety systems for cooling the core after emergency shut down. The reactor fuel rods are occasionally replaced by moving them from the reactor pressure vessel to the spent fuel pool. A typical fuel cycle lasts 18—24 months, with about one third of fuel assemblies being replaced during a refueling outage.

The remaining fuel assemblies are shuffled to new core locations to maximize the efficiency and power produced in the next fuel cycle. Because they are hot both radioactively and thermally, this is done via cranes and under water. For this reason the spent fuel storage pools are above the reactor in typical installations. They are shielded by water several times their height, and stored in rigid arrays in which their geometry is controlled to avoid criticality.

In the Fukushima Daiichi nuclear disaster this became problematic because water was lost as it was heated by the spent fuel from one or more spent fuel pools and the earthquake could have altered the geometry. Normally the fuel rods are kept sufficiently cool in the reactor and spent fuel pools that this is not a concern, and the cladding remains intact for the life of the rod.

Research into nuclear power in the US was led by the three military services. The Navy, seeing the possibility of turning submarines into full-time underwater vehicles, and ships that could steam around the world without refueling, sent their man in engineering, Captain Hyman Rickover to run their nuclear power program. Rickover decided on the PWR route for the Navy, as the early researchers in the field of nuclear power feared that the direct production of steam within a reactor would cause instability, while they knew that the use of pressurized water would definitively work as a means of heat transfer.

This concern led to the US\’s first research effort in nuclear power being devoted to the PWR, which was highly suited for naval vessels submarines, especially , as space was at a premium, and PWRs could be made compact and high-power enough to fit into such vessels. But other researchers wanted to investigate whether the supposed instability caused by boiling water in a reactor core would really cause instability.

During early reactor development, a small group of engineers accidentally increased the reactor power level on an experimental reactor to such an extent that the water quickly boiled. This shut down the reactor, indicating the useful self-moderating property in emergency circumstances. In particular, Samuel Untermyer II , a researcher at Argonne National Laboratory , proposed and oversaw a series of experiments: the BORAX experiments —to see if a boiling water reactor would be feasible for use in energy production.

He found that it was, after subjecting his reactors to quite strenuous tests, proving the safety principles of the BWR. Following this series of tests, GE got involved and collaborated with Argonne National Laboratory [7] to bring this technology to market. The literature does not indicate why this was the case, but it was eliminated on production models of the BWR.

The first generation of production boiling water reactors saw the incremental development of the unique and distinctive features of the BWR: 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, the distinctive general layout of the reactor building, and the standardization of reactor control and safety systems.

The vast majority of BWRs in service throughout the world belong to one of these design phases. Containment variants were constructed using either concrete or steel for the Primary Containment, Drywell and Wetwell in various combinations. See List of boiling water reactors. The ABWR was developed in the late s and early s, 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 MWe per reactor , and a significantly lowered probability of core damage.

Most significantly, the ABWR was a completely standardized design, that could be made for series production. This smaller megawatt electrical reactor was notable for its incorporation—for the first time ever in a light water reactor [ citation needed ] —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 solely through operation of natural forces if a safety-related contingency developed.

For example, if the reactor got too hot, it would trigger a system that would release soluble neutron absorbers generally a solution of borated materials, or a solution of borax , or materials that greatly hamper a chain reaction by absorbing neutrons, into the reactor core. The tank containing the soluble neutron absorbers would be located above the reactor, and the absorption solution, once the system was triggered, would flow into the core through force of gravity, and bring the reaction to a near-complete stop.

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 refueling 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 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 submitted [ when? During a period beginning in the late s, 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 1, MWe 4, MWth.

Reactor start up criticality is achieved by withdrawing control rods from the core to raise core reactivity to a level where it is evident that the nuclear chain reaction is self-sustaining. This is known as \”going critical\”. Control rod withdrawal is performed slowly, as to carefully monitor core conditions as the reactor approaches criticality. When the reactor is observed to become slightly super-critical, that is, reactor power is increasing on its own, the reactor is declared critical. Rod motion is performed using rod drive control systems.

This allows a reactor operator to evenly increase the core\’s reactivity until the reactor is critical. Older BWR designs use a manual control system, which is usually limited to controlling one or four control rods at a time, and only through a series of notched positions with fixed intervals between these positions. Due to the limitations of the manual control system, it is possible while starting-up that the core can be placed into a condition where movement of a single control rod can cause a large nonlinear reactivity change, which could heat fuel elements to the point they fail melt, ignite, weaken, etc.

As a result, GE developed a set of rules in called BPWS Banked Position Withdrawal Sequence which help minimize the effect of any single control rod movement and prevent fuel damage in the case of a control rod drop accident.

Then, either all of the A control rods or B control rods are pulled full out in a defined sequence to create a \” checkerboard \” pattern. Next, the opposing group B or A is pulled in a defined sequence to positions 02, then 04, 08, 16, and finally full out 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. Nuclear fuel could be damaged by film boiling; this would cause the fuel cladding to overheat and fail. 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.

Experimental data is conservatively applied to BWR fuel to ensure that the transition to film boiling does not occur during normal or transient operation. 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. 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 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.

APLHGR, being an average of the Linear Heat Generation Rate LHGR , a measure of the decay heat present in the fuel bundles, is a margin of safety associated with the potential for fuel failure to occur during a LBLOCA large-break loss-of-coolant accident — a massive pipe rupture leading to catastrophic loss of coolant pressure within the reactor, considered the most threatening \”design basis accident\” in probabilistic risk assessment and nuclear safety and security , which is anticipated to lead to the temporary exposure of the core; this core drying-out event is termed core \”uncovery\”, for the core loses its heat-removing cover of coolant, in the case of a BWR, light water.

BWR designs incorporate failsafe protection systems to rapidly cool and make safe the uncovered fuel prior to it reaching this temperature; these failsafe systems are known as the Emergency Core Cooling System. The ECCS is designed to rapidly flood the reactor pressure vessel, spray water on the core itself, and sufficiently cool the reactor fuel in this event.

However, like any system, the ECCS has limits, in this case, to its cooling capacity, and there is a possibility that fuel could be designed that produces so much decay heat that the ECCS would be overwhelmed and could not cool it down successfully. So as to prevent this from happening, it is required that the decay heat stored in the fuel assemblies at any one time does not overwhelm the ECCS.

APLHGR is monitored to ensure that the reactor is not operated at an average power level that would defeat the primary containment systems. Their approach is to simulate worst case events when the reactor is in its most vulnerable state. During the first nuclear heatup, nuclear fuel pellets can crack. The jagged edges of the pellet can rub and interact with the inner cladding wall.

A boiling water reactor BWR is a type of light water 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 , which is also a type of light water nuclear reactor.

In a PWR, the reactor core heats water, which does not boil. This hot water then exchanges heat with a lower pressure system, which turns water into steam that drives the turbine.

The main present manufacturer is GE Hitachi Nuclear Energy , which specializes in the design and construction of this type of reactor. A boiling water reactor uses demineralized 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 it 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 atm 7. In comparison, there is no significant boiling allowed in a pressurized water reactor PWR because of the high pressure maintained in its primary loop—approximately atm 16 MPa, psi. Steam exiting 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 or annulus region and combines with water exiting the moisture separators. The feedwater subcools the saturated water from the moisture separators. This water now flows down the downcomer or annulus 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 degree turn and moves up through the lower core plate into the nuclear core, where the fuel elements heat the water. 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, and allows achieving higher power levels that would not otherwise be possible.

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 moisture 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 or annulus region.

In the downcomer or annulus 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 \”wet\” steam goes through a tortuous path where the water droplets are slowed and directed out into the downcomer or annulus region. The \”dry\” steam then exits the RPV through four main steam lines and goes to the turbine.

Reactor power is controlled via two methods: by inserting or withdrawing control rods control blades 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.

Differently from the PWR, in a BWR the control rods boron carbide plates are inserted from below to give a more homogeneous distribution of the power: in the upper side the density of the water is lower due to vapour formation, making the neutron moderation less efficient and the fission probability lower. In normal operation, the control rods are only used to keep a homogeneous power distribution in the reactor and to compensate for the consumption of the fuel, while the power is controlled through the water flow see below.

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 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 enough to be absorbed by the fuel, and reactor power decreases.

Reactor pressure in a BWR is controlled by the main turbine or main steam bypass valves. Unlike a PWR, where the turbine steam demand is set manually by the operators, in a BWR, the turbine valves will modulate to maintain reactor pressure at a setpoint. Under this control mode, the turbine output will automatically follow reactor power changes. These bypass valves will automatically or manually modulate as necessary to maintain reactor pressure and control the reactor\’s heatup and cooldown rates while steaming is still in progress.

Reactor water level is controlled by the main feedwater system. From about 0. At low power conditions, the feedwater controller acts as a simple PID control by watching reactor water level. At high power conditions, the controller is switched to a \”Three-Element\” control mode, where the controller looks at the current water level in the reactor, as well as the amount of water going in and the amount of steam leaving the reactor.

By using the water injection and steam flow rates, the feed water control system can rapidly anticipate water level deviations and respond to maintain water level within a few inches of set point. At this power level a single feedwater pump can maintain the core water level.

If all feedwater is lost, the reactor will scram and the Emergency Core Cooling System is used to restore reactor water level. 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 due to neutron capture from the water, 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, with a 7-second half-life , so the turbine hall can be entered soon after the reactor is shut down.

BWR steam turbines employ a high-pressure turbine designed to handle saturated steam, and multiple low-pressure turbines. The high-pressure turbine receives steam directly from the reactor. The high-pressure turbine exhaust passes through a steam reheater which superheats the steam to over degrees F for the low-pressure turbines to use.

The exhaust of the low-pressure turbines is sent to the main condenser. The steam reheaters take some of the turbine\’s steam and use it as a heating source to reheat what comes out of the high-pressure turbine exhaust. While the reheaters take steam away from the turbine, the net result is that the reheaters improve the thermodynamic efficiency of the plant.

A modern BWR fuel assembly comprises 74 to fuel rods , and there are up to approximately assemblies in a reactor core , holding up to approximately short tons of low-enriched 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.

A modern reactor has many safety systems that are designed with a defence in depth philosophy, which is a design philosophy that is integrated throughout construction and commissioning. A BWR is similar to a pressurized water reactor PWR in that the reactor will continue to produce heat even after the fission reactions have stopped, which could make a core damage incident possible.

This heat is produced by the radioactive decay of fission products and materials that have been activated by neutron absorption. BWRs contain multiple safety systems for cooling the core after emergency shut down. The reactor fuel rods are occasionally replaced by moving them from the reactor pressure vessel to the spent fuel pool.

A typical fuel cycle lasts 18—24 months, with about one third of fuel assemblies being replaced during a refueling outage. The remaining fuel assemblies are shuffled to new core locations to maximize the efficiency and power produced in the next fuel cycle.

Because they are hot both radioactively and thermally, this is done via cranes and under water. For this reason the spent fuel storage pools are above the reactor in typical installations. They are shielded by water several times their height, and stored in rigid arrays in which their geometry is controlled to avoid criticality. In the Fukushima Daiichi nuclear disaster this became problematic because water was lost as it was heated by the spent fuel from one or more spent fuel pools and the earthquake could have altered the geometry.

Normally the fuel rods are kept sufficiently cool in the reactor and spent fuel pools that this is not a concern, and the cladding remains intact for the life of the rod. Research into nuclear power in the US was led by the three military services. The Navy, seeing the possibility of turning submarines into full-time underwater vehicles, and ships that could steam around the world without refueling, sent their man in engineering, Captain Hyman Rickover to run their nuclear power program.

Rickover decided on the PWR route for the Navy, as the early researchers in the field of nuclear power feared that the direct production of steam within a reactor would cause instability, while they knew that the use of pressurized water would definitively work as a means of heat transfer.

This concern led to the US\’s first research effort in nuclear power being devoted to the PWR, which was highly suited for naval vessels submarines, especially , as space was at a premium, and PWRs could be made compact and high-power enough to fit into such vessels. But other researchers wanted to investigate whether the supposed instability caused by boiling water in a reactor core would really cause instability.

During early reactor development, a small group of engineers accidentally increased the reactor power level on an experimental reactor to such an extent that the water quickly boiled.

This shut down the reactor, indicating the useful self-moderating property in emergency circumstances. In particular, Samuel Untermyer II , a researcher at Argonne National Laboratory , proposed and oversaw a series of experiments: the BORAX experiments —to see if a boiling water reactor would be feasible for use in energy production.

He found that it was, after subjecting his reactors to quite strenuous tests, proving the safety principles of the BWR. Following this series of tests, GE got involved and collaborated with Argonne National Laboratory [7] to bring this technology to market.

The literature does not indicate why this was the case, but it was eliminated on production models of the BWR. The first generation of production boiling water reactors saw the incremental development of the unique and distinctive features of the BWR: 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, the distinctive general layout of the reactor building, and the standardization of reactor control and safety systems.

The vast majority of BWRs in service throughout the world belong to one of these design phases. Containment variants were constructed using either concrete or steel for the Primary Containment, Drywell and Wetwell in various combinations. See List of boiling water reactors. The ABWR was developed in the late s and early s, 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 MWe per reactor , and a significantly lowered probability of core damage.

Most significantly, the ABWR was a completely standardized design, that could be made for series production. This smaller megawatt electrical reactor was notable for its incorporation—for the first time ever in a light water reactor [ citation needed ] —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 solely through operation of natural forces if a safety-related contingency developed.

For example, if the reactor got too hot, it would trigger a system that would release soluble neutron absorbers generally a solution of borated materials, or a solution of borax , or materials that greatly hamper a chain reaction by absorbing neutrons, into the reactor core.

The tank containing the soluble neutron absorbers would be located above the reactor, and the absorption solution, once the system was triggered, would flow into the core through force of gravity, and bring the reaction to a near-complete stop.

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 refueling 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 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.

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Nuclear reactor.Reaktor 6 manual german free

 
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Человек-гигант удивленно поднял брови. Даже перепачканная сажей и промокшая, Сьюзан Флетчер производила более сильное впечатление, чем он мог предположить. – А gwrman спросил. Бринкерхофф покачал головой.

Но… офицер ничего не сказал о… – Разумеется. Я не сказал ему про спутницу.  – Взмахом руки Клушар величественно отверг вопрос Беккера.

Some of these methods arise naturally from the physics of radioactive decay and are simply accounted for during the reactor\’s operation, while others are mechanisms engineered into the reactor design for a distinct purpose.

The fastest method for adjusting levels of fission-inducing neutrons in a reactor is via movement of the control rods. Control rods are made of neutron poisons and therefore absorb neutrons.

When a control rod is inserted deeper into the reactor, it absorbs more neutrons than the material it displaces — often the moderator. This action results in fewer neutrons available to cause fission and reduces the reactor\’s power output.

Conversely, extracting the control rod will result in an increase in the rate of fission events and an increase in power. The physics of radioactive decay also affects neutron populations in a reactor. One such process is delayed neutron emission by a number of neutron-rich fission isotopes. These delayed neutrons account for about 0. The fission products which produce delayed neutrons have half-lives for their decay by neutron emission that range from milliseconds to as long as several minutes, and so considerable time is required to determine exactly when a reactor reaches the critical point.

Keeping the reactor in the zone of chain reactivity where delayed neutrons are necessary to achieve a critical mass state allows mechanical devices or human operators to control a chain reaction in \”real time\”; otherwise the time between achievement of criticality and nuclear meltdown as a result of an exponential power surge from the normal nuclear chain reaction, would be too short to allow for intervention.

This last stage, where delayed neutrons are no longer required to maintain criticality, is known as the prompt critical point. There is a scale for describing criticality in numerical form, in which bare criticality is known as zero dollars and the prompt critical point is one dollar , and other points in the process interpolated in cents.

In some reactors, the coolant also acts as a neutron moderator. A moderator increases the power of the reactor by causing the fast neutrons that are released from fission to lose energy and become thermal neutrons.

Thermal neutrons are more likely than fast neutrons to cause fission. A higher temperature coolant would be less dense, and therefore a less effective moderator. In other reactors the coolant acts as a poison by absorbing neutrons in the same way that the control rods do.

In these reactors power output can be increased by heating the coolant, which makes it a less dense poison. Nuclear reactors generally have automatic and manual systems to scram the reactor in an emergency shut down. These systems insert large amounts of poison often boron in the form of boric acid into the reactor to shut the fission reaction down if unsafe conditions are detected or anticipated.

Most types of reactors are sensitive to a process variously known as xenon poisoning, or the iodine pit.

The common fission product Xenon produced in the fission process acts as a neutron poison that absorbs neutrons and therefore tends to shut the reactor down.

Xenon accumulation can be controlled by keeping power levels high enough to destroy it by neutron absorption as fast as it is produced. Fission also produces iodine , which in turn decays with a half-life of 6. When the reactor is shut down, iodine continues to decay to xenon, making restarting the reactor more difficult for a day or two, as the xenon decays into cesium, which is not nearly as poisonous as xenon, with a half-life of 9. This temporary state is the \”iodine pit.

As the extra xenon is transmuted to xenon, which is much less a neutron poison, within a few hours the reactor experiences a \”xenon burnoff power transient\”.

Control rods must be further inserted to replace the neutron absorption of the lost xenon Failure to properly follow such a procedure was a key step in the Chernobyl disaster.

Reactors used in nuclear marine propulsion especially nuclear submarines often cannot be run at continuous power around the clock in the same way that land-based power reactors are normally run, and in addition often need to have a very long core life without refueling.

For this reason many designs use highly enriched uranium but incorporate burnable neutron poison in the fuel rods. The energy released in the fission process generates heat, some of which can be converted into usable energy. A common method of harnessing this thermal energy is to use it to boil water to produce pressurized steam which will then drive a steam turbine that turns an alternator and generates electricity. The neutron was discovered in by British physicist James Chadwick.

He filed a patent for his idea of a simple reactor the following year while working at the Admiralty in London. Inspiration for a new type of reactor using uranium came from the discovery by Lise Meitner , Fritz Strassmann and Otto Hahn in that bombardment of uranium with neutrons provided by an alpha-on-beryllium fusion reaction, a \” neutron howitzer \” produced a barium residue, which they reasoned was created by the fissioning of the uranium nuclei.

The U. The following year the U. Government received the Frisch—Peierls memorandum from the UK, which stated that the amount of uranium needed for a chain reaction was far lower than had previously been thought. Eventually, the first artificial nuclear reactor, Chicago Pile-1 , was constructed at the University of Chicago , by a team led by Italian physicist Enrico Fermi , in late By this time, the program had been pressured for a year by U.

The reactor support structure was made of wood, which supported a pile hence the name of graphite blocks, embedded in which was natural uranium oxide \’pseudospheres\’ or \’briquettes\’. Soon after the Chicago Pile, the U. The primary purpose for the largest reactors located at the Hanford Site in Washington , was the mass production of plutonium for nuclear weapons. Fermi and Szilard applied for a patent on reactors on 19 December Its issuance was delayed for 10 years because of wartime secrecy.

Atomic Energy Commission produced 0. Besides the military uses of nuclear reactors, there were political reasons to pursue civilian use of atomic energy. This diplomacy led to the dissemination of reactor technology to U. It produced around 5 MW electrical. It was built after the F-1 nuclear reactor which was the first reactor to go critical in Europe, and was also built by the Soviet Union. All commercial power reactors are based on nuclear fission. They generally use uranium and its product plutonium as nuclear fuel , though a thorium fuel cycle is also possible.

Fission reactors can be divided roughly into two classes, depending on the energy of the neutrons that sustain the fission chain reaction :. In principle, fusion power could be produced by nuclear fusion of elements such as the deuterium isotope of hydrogen. While an ongoing rich research topic since at least the s, no self-sustaining fusion reactor for any purpose has ever been built. More than a dozen advanced reactor designs are in various stages of development. Rolls-Royce aims to sell nuclear reactors for the production of synfuel for aircraft.

Generation IV reactors are a set of theoretical nuclear reactor designs currently being researched. These designs are generally not expected to be available for commercial construction before Current reactors in operation around the world are generally considered second- or third-generation systems, with the first-generation systems having been retired some time ago.

The primary goals being to improve nuclear safety, improve proliferation resistance, minimize waste and natural resource utilization, and to decrease the cost to build and run such plants. Generation V reactors are designs which are theoretically possible, but which are not being actively considered or researched at present.

Though some generation V reactors could potentially be built with current or near term technology, they trigger little interest for reasons of economics, practicality, or safety.

Controlled nuclear fusion could in principle be used in fusion power plants to produce power without the complexities of handling actinides , but significant scientific and technical obstacles remain.

Several fusion reactors have been built, but reactors have never been able to release more energy than the amount of energy used in the process. Despite research having started in the s, no commercial fusion reactor is expected before The ITER project is currently leading the effort to harness fusion power.

Thermal reactors generally depend on refined and enriched uranium. Some nuclear reactors can operate with a mixture of plutonium and uranium see MOX. The process by which uranium ore is mined, processed, enriched, used, possibly reprocessed and disposed of is known as the nuclear fuel cycle. Enrichment involves increasing the percentage of U and is usually done by means of gaseous diffusion or gas centrifuge.

The enriched result is then converted into uranium dioxide powder, which is pressed and fired into pellet form. These pellets are stacked into tubes which are then sealed and called fuel rods. Many of these fuel rods are used in each nuclear reactor. Theft risk of this fuel potentially used in the production of a nuclear weapon has led to campaigns advocating conversion of this type of reactor to low-enrichment uranium which poses less threat of proliferation.

Fissile U and non-fissile but fissionable and fertile U are both used in the fission process. U is fissionable by thermal i. A thermal neutron is one which is moving about the same speed as the atoms around it. Since all atoms vibrate proportionally to their absolute temperature, a thermal neutron has the best opportunity to fission U when it is moving at this same vibrational speed.

On the other hand, U is more likely to capture a neutron when the neutron is moving very fast. This U atom will soon decay into plutonium, which is another fuel. Pu is a viable fuel and must be accounted for even when a highly enriched uranium fuel is used. Plutonium fissions will dominate the U fissions in some reactors, especially after the initial loading of U is spent.

Plutonium is fissionable with both fast and thermal neutrons, which make it ideal for either nuclear reactors or nuclear bombs. Most reactor designs in existence are thermal reactors and typically use water as a neutron moderator moderator means that it slows down the neutron to a thermal speed and as a coolant.

But in a fast breeder reactor , some other kind of coolant is used which will not moderate or slow the neutrons down much. This enables fast neutrons to dominate, which can effectively be used to constantly replenish the fuel supply.

By merely placing cheap unenriched uranium into such a core, the non-fissionable U will be turned into Pu, \”breeding\” fuel. In thorium fuel cycle thorium absorbs a neutron in either a fast or thermal reactor. The thorium beta decays to protactinium and then to uranium , which in turn is used as fuel. Hence, like uranium , thorium is a fertile material.

The amount of energy in the reservoir of nuclear fuel is frequently expressed in terms of \”full-power days,\” which is the number of hour periods days a reactor is scheduled for operation at full power output for the generation of heat energy. The number of full-power days in a reactor\’s operating cycle between refueling outage times is related to the amount of fissile uranium U contained in the fuel assemblies at the beginning of the cycle.

A higher percentage of U in the core at the beginning of a cycle will permit the reactor to be run for a greater number of full-power days. At the end of the operating cycle, the fuel in some of the assemblies is \”spent\”, having spent four to six years in the reactor producing power.

This spent fuel is discharged and replaced with new fresh fuel assemblies. Plants typically operate on 18 month refueling cycles, or 24 month refueling cycles. This means that one refueling, replacing only one-third of the fuel, can keep a nuclear reactor at full power for nearly two years.

This nuclear waste is highly radioactive and its toxicity presents a danger for thousands of years. The spent fuel pool is a large pool of water that provides cooling and shielding of the spent nuclear fuel.

After loading into dry shielded casks, the casks are stored on-site in a specially guarded facility in impervious concrete bunkers. On-site fuel storage facilities are designed to withstand the impact of commercial airliners, with little to no damage to the spent fuel. An average on-site fuel storage facility can hold 30 years of spent fuel in a space smaller than a football field. In a CANDU reactor, this also allows individual fuel elements to be situated within the reactor core that are best suited to the amount of U in the fuel element.

The amount of energy extracted from nuclear fuel is called its burnup , which is expressed in terms of the heat energy produced per initial unit of fuel weight. Burnup is commonly expressed as megawatt days thermal per metric ton of initial heavy metal.

Nuclear safety covers the actions taken to prevent nuclear and radiation accidents and incidents or to limit their consequences. The nuclear power industry has improved the safety and performance of reactors, and has proposed new, safer but generally untested reactor designs but there is no guarantee that the reactors will be designed, built and operated correctly.

Serious, though rare, nuclear and radiation accidents have occurred. Nuclear reactors have been launched into Earth orbit at least 34 times. A number of incidents connected with the unmanned nuclear-reactor-powered Soviet RORSAT especially Kosmos radar satellite which resulted in nuclear fuel reentering the Earth\’s atmosphere from orbit and being dispersed in northern Canada January Almost two billion years ago a series of self-sustaining nuclear fission \”reactors\” self-assembled in the area now known as Oklo in Gabon , West Africa.

The conditions at that place and time allowed a natural nuclear fission to occur with circumstances that are similar to the conditions in a constructed nuclear reactor. First discovered in by French physicist Francis Perrin , they are collectively known as the Oklo Fossil Reactors.

Self-sustaining nuclear fission reactions took place in these reactors approximately 1. Such reactors can no longer form on Earth in its present geologic period. Radioactive decay of formerly more abundant uranium over the time span of hundreds of millions of years has reduced the proportion of this naturally occurring fissile isotope to below the amount required to sustain a chain reaction with only plain water as a moderator.

The natural nuclear reactors formed when a uranium-rich mineral deposit became inundated with groundwater that acted as a neutron moderator, and a strong chain reaction took place. The water moderator would boil away as the reaction increased, slowing it back down again and preventing a meltdown. The fission reaction was sustained for hundreds of thousands of years, cycling on the order of hours to a few days.

These natural reactors are extensively studied by scientists interested in geologic radioactive waste disposal. They offer a case study of how radioactive isotopes migrate through the Earth\’s crust. This is a significant area of controversy as opponents of geologic waste disposal fear that isotopes from stored waste could end up in water supplies or be carried into the environment. Nuclear reactors produce tritium as part of normal operations, which is eventually released into the environment in trace quantities.

As an isotope of hydrogen , tritium T frequently binds to oxygen and forms T 2 O. This molecule is chemically identical to H 2 O and so is both colorless and odorless, however the additional neutrons in the hydrogen nuclei cause the tritium to undergo beta decay with a half-life of Despite being measurable, the tritium released by nuclear power plants is minimal.

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, with a 7-second half-life , so the turbine hall can be entered soon after the reactor is shut down.

BWR steam turbines employ a high-pressure turbine designed to handle saturated steam, and multiple low-pressure turbines. The high-pressure turbine receives steam directly from the reactor.

The high-pressure turbine exhaust passes through a steam reheater which superheats the steam to over degrees F for the low-pressure turbines to use. The exhaust of the low-pressure turbines is sent to the main condenser. The steam reheaters take some of the turbine\’s steam and use it as a heating source to reheat what comes out of the high-pressure turbine exhaust. While the reheaters take steam away from the turbine, the net result is that the reheaters improve the thermodynamic efficiency of the plant.

A modern BWR fuel assembly comprises 74 to fuel rods , and there are up to approximately assemblies in a reactor core , holding up to approximately short tons of low-enriched 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.

A modern reactor has many safety systems that are designed with a defence in depth philosophy, which is a design philosophy that is integrated throughout construction and commissioning. A BWR is similar to a pressurized water reactor PWR in that the reactor will continue to produce heat even after the fission reactions have stopped, which could make a core damage incident possible. This heat is produced by the radioactive decay of fission products and materials that have been activated by neutron absorption.

BWRs contain multiple safety systems for cooling the core after emergency shut down. The reactor fuel rods are occasionally replaced by moving them from the reactor pressure vessel to the spent fuel pool. A typical fuel cycle lasts 18—24 months, with about one third of fuel assemblies being replaced during a refueling outage. The remaining fuel assemblies are shuffled to new core locations to maximize the efficiency and power produced in the next fuel cycle.

Because they are hot both radioactively and thermally, this is done via cranes and under water. For this reason the spent fuel storage pools are above the reactor in typical installations.

They are shielded by water several times their height, and stored in rigid arrays in which their geometry is controlled to avoid criticality.

In the Fukushima Daiichi nuclear disaster this became problematic because water was lost as it was heated by the spent fuel from one or more spent fuel pools and the earthquake could have altered the geometry. Normally the fuel rods are kept sufficiently cool in the reactor and spent fuel pools that this is not a concern, and the cladding remains intact for the life of the rod. Research into nuclear power in the US was led by the three military services.

The Navy, seeing the possibility of turning submarines into full-time underwater vehicles, and ships that could steam around the world without refueling, sent their man in engineering, Captain Hyman Rickover to run their nuclear power program. Rickover decided on the PWR route for the Navy, as the early researchers in the field of nuclear power feared that the direct production of steam within a reactor would cause instability, while they knew that the use of pressurized water would definitively work as a means of heat transfer.

This concern led to the US\’s first research effort in nuclear power being devoted to the PWR, which was highly suited for naval vessels submarines, especially , as space was at a premium, and PWRs could be made compact and high-power enough to fit into such vessels. But other researchers wanted to investigate whether the supposed instability caused by boiling water in a reactor core would really cause instability.

During early reactor development, a small group of engineers accidentally increased the reactor power level on an experimental reactor to such an extent that the water quickly boiled.

This shut down the reactor, indicating the useful self-moderating property in emergency circumstances. In particular, Samuel Untermyer II , a researcher at Argonne National Laboratory , proposed and oversaw a series of experiments: the BORAX experiments —to see if a boiling water reactor would be feasible for use in energy production.

He found that it was, after subjecting his reactors to quite strenuous tests, proving the safety principles of the BWR. Following this series of tests, GE got involved and collaborated with Argonne National Laboratory [7] to bring this technology to market. The literature does not indicate why this was the case, but it was eliminated on production models of the BWR. The first generation of production boiling water reactors saw the incremental development of the unique and distinctive features of the BWR: 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, the distinctive general layout of the reactor building, and the standardization of reactor control and safety systems.

The vast majority of BWRs in service throughout the world belong to one of these design phases. Containment variants were constructed using either concrete or steel for the Primary Containment, Drywell and Wetwell in various combinations. See List of boiling water reactors. The ABWR was developed in the late s and early s, 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 MWe per reactor , and a significantly lowered probability of core damage.

Most significantly, the ABWR was a completely standardized design, that could be made for series production. This smaller megawatt electrical reactor was notable for its incorporation—for the first time ever in a light water reactor [ citation needed ] —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 solely through operation of natural forces if a safety-related contingency developed.

For example, if the reactor got too hot, it would trigger a system that would release soluble neutron absorbers generally a solution of borated materials, or a solution of borax , or materials that greatly hamper a chain reaction by absorbing neutrons, into the reactor core.

The tank containing the soluble neutron absorbers would be located above the reactor, and the absorption solution, once the system was triggered, would flow into the core through force of gravity, and bring the reaction to a near-complete stop. 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 refueling 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 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 submitted [ when? During a period beginning in the late s, 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 1, MWe 4, MWth.

Reactor start up criticality is achieved by withdrawing control rods from the core to raise core reactivity to a level where it is evident that the nuclear chain reaction is self-sustaining. This is known as \”going critical\”. Control rod withdrawal is performed slowly, as to carefully monitor core conditions as the reactor approaches criticality.

When the reactor is observed to become slightly super-critical, that is, reactor power is increasing on its own, the reactor is declared critical. Rod motion is performed using rod drive control systems. This allows a reactor operator to evenly increase the core\’s reactivity until the reactor is critical. Older BWR designs use a manual control system, which is usually limited to controlling one or four control rods at a time, and only through a series of notched positions with fixed intervals between these positions.

Due to the limitations of the manual control system, it is possible while starting-up that the core can be placed into a condition where movement of a single control rod can cause a large nonlinear reactivity change, which could heat fuel elements to the point they fail melt, ignite, weaken, etc. As a result, GE developed a set of rules in called BPWS Banked Position Withdrawal Sequence which help minimize the effect of any single control rod movement and prevent fuel damage in the case of a control rod drop accident.

Then, either all of the A control rods or B control rods are pulled full out in a defined sequence to create a \” checkerboard \” pattern. Next, the opposing group B or A is pulled in a defined sequence to positions 02, then 04, 08, 16, and finally full out 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.

Nuclear fuel could be damaged by film boiling; this would cause the fuel cladding to overheat and fail. 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. Experimental data is conservatively applied to BWR fuel to ensure that the transition to film boiling does not occur during normal or transient operation. 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. 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 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. APLHGR, being an average of the Linear Heat Generation Rate LHGR , a measure of the decay heat present in the fuel bundles, is a margin of safety associated with the potential for fuel failure to occur during a LBLOCA large-break loss-of-coolant accident — a massive pipe rupture leading to catastrophic loss of coolant pressure within the reactor, considered the most threatening \”design basis accident\” in probabilistic risk assessment and nuclear safety and security , which is anticipated to lead to the temporary exposure of the core; this core drying-out event is termed core \”uncovery\”, for the core loses its heat-removing cover of coolant, in the case of a BWR, light water.

BWR designs incorporate failsafe protection systems to rapidly cool and make safe the uncovered fuel prior to it reaching this temperature; these failsafe systems are known as the Emergency Core Cooling System.

The ECCS is designed to rapidly flood the reactor pressure vessel, spray water on the core itself, and sufficiently cool the reactor fuel in this event. However, like any system, the ECCS has limits, in this case, to its cooling capacity, and there is a possibility that fuel could be designed that produces so much decay heat that the ECCS would be overwhelmed and could not cool it down successfully. So as to prevent this from happening, it is required that the decay heat stored in the fuel assemblies at any one time does not overwhelm the ECCS.

APLHGR is monitored to ensure that the reactor is not operated at an average power level that would defeat the primary containment systems. Their approach is to simulate worst case events when the reactor is in its most vulnerable state.

During the first nuclear heatup, nuclear fuel pellets can crack. The jagged edges of the pellet can rub and interact with the inner cladding wall. During power increases in the fuel pellet, the ceramic fuel material expands faster than the fuel cladding, and the jagged edges of the fuel pellet begin to press into the cladding, potentially causing a perforation. To prevent this from occurring, two corrective actions were taken. The first is the inclusion of a thin barrier layer against the inner walls of the fuel cladding which are resistant to perforation due to pellet-clad interactions, and the second is a set of rules created under PCIOMR.

This means, for the first nuclear heatup of each fuel element, that local bundle power must be ramped very slowly to prevent cracking of the fuel pellets and limit the differences in the rates of thermal expansion of the fuel.

PCIOMR analysis look at local power peaks and xenon transients which could be caused by control rod position changes or rapid power changes to ensure that local power rates never exceed maximum ratings. From Wikipedia, the free encyclopedia. Type of nuclear reactor that directly boils water.

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