Fugen Nuclear Power Plant
Fugen ふげん was a prototype Japanese nuclear test reactor. Fugen was a domestic Japanese design for a demonstration Advanced Thermal Reactor, it was a heavy water moderated, boiling light water cooled reactor. The reactor was started in 1979 and shut down in 2003; as of 2018, it is undergoing decommissioning. It is located in the city of Tsuruga, Fukui; the name "Fugen" is derived from a Buddhist deity. The reactor was the first in the world to use a full MOX fuel core, it had the most in the world. It has received the title of a historic landmark from the American Nuclear Society; the design boils ordinary water like a boiling water reactor but uses heavy water as a moderator as in a CANDU reactor. The electrical output was 165 MW and the thermal output was 557 MW. Core temperature: 300 °C Pellet centerline temperature: 2200 °C Fuel conversion time: 6 monthsThe plant is located on a site that covers 267,694 m2, it employed 256 workers. 14–16 April 1997: A tritium leakage was announced to the responsible authorities 30 hours after the event.
During the following investigation it was shown that it had 11 similar incidents. Five managers of the operator at that time resigned. 8 April 2002: About 200 cubic meters of steam escaped from a defective pipe. The reactor was switched off. During dismantling operations it was found that walls with controls did not have the necessary strength at 25 of 34 points. Official website Fugen Decommissioning Engineering Center Transcript of NHK TV special in Japanese NHK video
Nuclear power in India
Nuclear power is the fifth-largest source of electricity in India after coal, gas and wind power. As of March 2018, India has 22 nuclear reactors in operation in 7 nuclear power plants, having a total installed capacity of 6,780 MW. Nuclear power produced a total of 35 TWh and supplied 3.22% of Indian electricity in 2017. 7 more reactors are under construction with a combined generation capacity of 4,300 MW. In October 2010, India drew up a plan to reach a nuclear power capacity of 63 GW in 2032, but after the 2011 Fukushima nuclear disaster in Japan people around proposed Indian nuclear power plant sites have launched protests, raising questions about atomic energy as a clean and safe alternative to fossil fuels. There have been mass protests against the French-backed 9,900 MW Jaitapur Nuclear Power Project in Maharashtra and the Russian-backed 2,000 MW Kudankulam Nuclear Power Plant in Tamil Nadu; the state government of West Bengal, has refused permission to a proposed 6,000 MW facility near the town of Haripur that intended to host six Russian reactors.
A Public Interest Litigation has been filed against the government’s civil nuclear programme at the Supreme Court. Nuclear power in India has suffered from low capacity factors; as of 2017, the lifetime weighted energy availability factor of the Indian fleet is 63.5%. However, capacity factors have been improving in recent years; the availability factor of Indian reactors was 69.4% in the years 2015-2017. One of the main reasons for the low capacity factors is lack of nuclear fuel. India has been making advances in the field of thorium-based fuels, working to design and develop a prototype for an atomic reactor using thorium and low-enriched uranium, a key part of India's three stage nuclear power programme; the country has recently re-initiated its involvement in the LENR research activities, in addition to supporting work done in the fusion power area through the ITER initiative. As early as 1901, the Geological Survey of India had recognised India as having significant deposits of radioactive ores, including pitchblende and thorianite.
In the ensuing 50 years, little to no effort was made to exploit those resources. During the 1920s and 1930s, Indian scientists maintained close links to their counterparts in Europe and the United States, were well aware of the latest developments in physics. Several Indian physicists, notably Daulat Singh Kothari, Meghnad Saha, Homi J. Bhabha and R. S. Krishnan, conducted pioneering research in nuclear physics in Europe during the 1930s. By 1939, Meghnad Saha, the Palit Professor of Physics at the University of Calcutta, had recognised the significance of the discovery of nuclear fission, had begun to conduct various experiments in his laboratory related to nuclear physics. In 1940, he incorporated nuclear physics into the university's post-graduate curriculum. In the same year, the Sir Dorabji Tata Trust sanctioned funds for installing a cyclotron at the University of Calcutta, but various difficulties related to the war delayed the project. In 1944, Homi J. Bhabha, a distinguished nuclear physicist who had established a research school at the Indian Institute of Science, wrote a letter to his distant cousin J. R. D. Tata, the chairman of the Tata Group.
He requested funds to establish a research institute of fundamental physics, "with special reference to cosmic rays and nuclear physics." The Tata Institute of Fundamental Research was inaugurated in Mumbai the following year. Following the atomic bombing of Hiroshima in August 1945, R. S. Krishnan, a nuclear physicist who had studied under Norman Feather and John Cockroft, who recognised the massive energy-generating potential of uranium, observed, "If the tremendous energy released from atomic explosions is made available to drive machinery, etc. it will bring about an industrial revolution of a far-reaching character." He further noted, the difficulties in harnessing nuclear power for peaceful usage, "...a great deal more research work is needed before atomic power can be put to industrial use."In March 1946, the Board of Scientific and Industrial Research, under the Council of Scientific and Industrial Research, set up an Atomic Research Committee under Bhabha's leadership to explore India's atomic energy resources and to suggest ways to develop and harness them, along with establishing contacts with similar organisations in other nations.
At the same time, the University of Travancore's research council met to discuss Travancore's future industrial development. Among other matters, the council made recommendations for developing the state's resources of monazite, a valuable thorium ore, ilmenite, with regard to their applications in atomic energy; the council suggested. This was followed by the deputation of Bhabha and Sir Shanti Swarup Bhatnagar, the Director of the CSIR, to Travancore in April 1947 and the establishment of a working relationship with the kingdom's dewan, Sir C. P. Ramaswami Iyer. Early in 1947, plans were made to establish a Uranium Unit under the Geological Survey of India, to focus on identifying and developing resources of uranium-bearing minerals. In June 1947, two months before Indian independence, Chakravarti Rajagopalachari Minister for Industry, Supply and Finance in the Interim Government of India, established an Advisory Board for Research in Atomic Energy. Chaired by Bhabha and placed under the CSIR, the Advisory Board included Saha and several other distinguished scientists, notably Sir K. S. Krishnan, the co-discoverer of the Raman effect, geologist Darashaw Nosherwan Wadia and Nazir Ahmed, a student of Ernest Rutherford.
A Joint Committee comprising the above scientists and th
Advanced boiling water reactor
The advanced boiling water reactor is a Generation III boiling water reactor. The ABWR is offered by GE Hitachi Nuclear Energy and Toshiba; the ABWR generates electrical power by using steam to power a turbine connected to a generator. Kashiwazaki-Kariwa unit 6 is considered the first Generation III reactor in the world. Boiling water reactors are the second most common form of light water reactor with a direct cycle design that uses fewer large steam supply components than the pressurized water reactor, which employs an indirect cycle; the ABWR is the present state of the art in boiling water reactors, is the first Generation III reactor design to be built, with several reactors complete and operating. The first reactors were built on time and under budget in Japan, with others under construction there and in Taiwan. ABWRs were on order including two reactors at the South Texas Project site; the projects in both Taiwan and US are both reported over-budgeted. The standard ABWR plant design has a net electrical output of about 1.35 GW, generated from about 3926 MW of thermal power.
The ABWR represents an evolutionary route for the BWR family, with numerous changes and improvements to previous BWR designs. Major areas of improvement include: The addition of reactor internal pumps mounted on the bottom of the reactor pressure vessel - 10 in total - which achieve improved performance while eliminating large recirculation pumps in containment and associated large-diameter and complex piping interfaces with the RPV. Only the RIP motor is located outside of the RPV in the ABWR. According to the Tier 1 Design Control Document, each RIP has a nominal capacity of 6912 m3/h; the control rod adjustment capabilities have been supplemented with the addition of an electro-hydraulic Fine Motion Control Rod Drive, allowing for fine position adjustment using an electrical motor, while not losing the reliability or redundancy of traditional hydraulic systems which are designed to accomplish rapid shutdown in 2.80 s from receipt of an initiating signal, or ARI in a greater but still insignificant time period.
The FMCRD improves defense-in-depth in the event of primary hydraulic and ARI contingencies. A digital Reactor Protection System ensures a high level of reliability and simplification for safety condition detection and response; this system initiates rapid hydraulic insertion of control rods for shutdown. Two-out-of-four per parameter rapid shutdown logic ensures that nuisance rapid shutdowns are not triggered by single instrument failures. RPS can trigger ARI, FMCRD rod run-in to shut down the nuclear chain reaction; the standby liquid control system actuation is provided as diverse logic in the unlikely event of an Anticipated Transient Without Scram. Digital reactor controls allow the control room to and control plant operations and processes. Separate redundant safety and non-safety related digital multiplexing buses allow for reliability and diversity of instrumentation and control. In particular, the reactor is automated for startup and for standard shutdown using automatic systems only.
Of course, human operators remain essential to reactor control and supervision, but much of the busy-work of bringing the reactor to power and descending from power can be automated at operator discretion. The Emergency Core Cooling System has been improved in many areas, providing a high level of defense-in-depth against accidents and incidents; the overall system has been divided up into 3 divisions. Previous BWRs had 2 divisions, uncovery was predicted to occur for a short time in the event of a severe accident, prior to ECCS response. Eighteen SORVs, eight of which are part of the ADS, ensure that RPV overpressure events are mitigated, that if necessary, that the reactor can be depressurized to a level where low pressure core flooder can be used. Further, LPCF can inject against much higher RPV pressures, providing an increased level of safety in the event of intermediate-sized breaks, which could be small enough to result in slow natural depressurization but could be large enough to result in high pressure corespray/coolant injection systems' capacities for response being overwhelmed by the size of the break.
Though the Class 1E power bus is still powered by 3 highly-reliable emergency diesel generators that are safety related, an additional Plant Investment Protection power bus using a combustion gas turbine is located on-site to generate electricity to provide defense-in-depth against station blackout contingencies as well as to power important but non-safety critical systems in the event of a loss of offsite power. Though one division of the ECCS does not have high pressu
A nuclear reactor known as an atomic pile, is a device used to initiate and control a self-sustained nuclear chain reaction. Nuclear reactors are used at nuclear power plants for electricity generation and in propulsion of ships. Heat from nuclear fission is passed to a working fluid; these either 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 of weapons-grade plutonium. Some are run only for research; as of early 2019, the IAEA reports there are 454 nuclear power reactors and 226 nuclear research reactors in operation around the world. Just as conventional power-stations generate electricity by harnessing the thermal energy released from burning fossil fuels, nuclear 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-235 or plutonium-239 absorbs a neutron, it may undergo nuclear fission.
The heavy nucleus splits into two or more lighter nuclei, releasing kinetic energy, gamma radiation, free neutrons. A portion of these neutrons may be absorbed by other fissile atoms and trigger further fission events, which release more neutrons, so on; this is known as a nuclear chain reaction. To control such a nuclear chain reaction, neutron poisons and neutron moderators can change the portion of neutrons that will go on to cause more fission. Nuclear reactors have automatic and manual systems to shut the fission reaction down if monitoring detects unsafe conditions. Used moderators include regular water, solid graphite and heavy water; some experimental types of reactor have used beryllium, hydrocarbons have been suggested as another possibility. The reactor core generates heat in a number of ways: The kinetic energy of fission products is converted to thermal energy when these nuclei collide with nearby atoms; the reactor absorbs some of the gamma rays produced during fission and converts their energy into heat.
Heat is produced by the radioactive decay of fission products and materials that have been activated by neutron absorption. This decay heat-source will remain for some time after the reactor is shut down. A kilogram of uranium-235 converted via nuclear processes releases three million times more energy than a kilogram of coal burned conventionally. A nuclear reactor coolant — water but sometimes a gas or a liquid metal or molten salt — is circulated past the reactor core to absorb the heat that it generates; the heat is carried away from the reactor and is used to generate steam. Most reactor systems employ a cooling system, physically separated from the water that will be boiled to produce pressurized steam for the turbines, like the pressurized water reactor. However, in some reactors the water for the steam turbines is boiled directly by the reactor core; 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 employ several methods of neutron control to adjust the reactor's power output. Some of these methods arising from the physics of radioactive decay and are 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 therefore tend to 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 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 0.65% of the total neutrons produced in fission, with the remainder released upon fission.
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, so considerable time is required to determine 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"; 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, other points in the process interpolated in cents. In some reactors, the coolant acts as a neutron moderator. A moderator increases the power of the reactor by causin
Pressurized water reactor
Pressurized water reactors constitute the large majority of the world's nuclear power plants and are one of three types of light water reactor, the other types being boiling water reactors and supercritical water reactors. In a PWR, the primary coolant is pumped under high pressure to the reactor core where it is heated by the energy released by the fission of atoms; the heated water flows to a steam generator where it transfers its thermal energy to a secondary system where steam is generated and flows to turbines which, in turn, spin an electric generator. In contrast to a boiling water reactor, pressure in the primary coolant loop prevents the water from boiling within the reactor. All LWRs use ordinary water as both neutron moderator. PWRs were designed to serve as nuclear marine propulsion for nuclear submarines and were used in the original design of the second commercial power plant at Shippingport Atomic Power Station. PWRs operating in the United States are considered Generation II reactors.
Russia's VVER reactors are similar to U. S. PWRs. France operates many PWRs to generate the bulk of its electricity. Several hundred PWRs are used for marine propulsion in aircraft carriers, nuclear submarines and ice breakers. In the US, they were designed at the Oak Ridge National Laboratory for use as a nuclear submarine power plant with a operational submarine power plant located at the Idaho National Engineering Lab. Follow-on work was conducted by Westinghouse Bettis Atomic Power Laboratory; the first purely commercial nuclear power plant at Shippingport Atomic Power Station was designed as a pressurized water reactor, on insistence from Admiral Hyman G. Rickover that a viable commercial plant would include none of the "crazy thermodynamic cycles that everyone else wants to build."The United States Army Nuclear Power Program operated pressurized water reactors from 1954 to 1974. Three Mile Island Nuclear Generating Station operated two pressurized water reactor plants, TMI-1 and TMI-2; the partial meltdown of TMI-2 in 1979 ended the growth in new construction of nuclear power plants in the United States for two decades.
The pressurized water reactor has three new Generation III reactor evolutionary designs: the AP-1000, VVER-1200, ACPR1000+, APR1400. Nuclear fuel in the reactor pressure vessel is engaged in a fission chain reaction, which produces heat, heating the water in the primary coolant loop by thermal conduction through the fuel cladding; the hot primary coolant is pumped into a heat exchanger called the steam generator, where it flows through hundreds or thousands of small tubes. Heat is transferred through the walls of these tubes to the lower pressure secondary coolant located on the sheet side of the exchanger where the coolant evaporates to pressurized steam; the transfer of heat is accomplished without mixing the two fluids to prevent the secondary coolant from becoming radioactive. Some common steam generator arrangements are single pass heat exchangers. In a nuclear power station, the pressurized steam is fed through a steam turbine which drives an electrical generator connected to the electric grid for transmission.
After passing through the turbine the secondary coolant is cooled condensed in a condenser. The condenser converts the steam to a liquid so that it can be pumped back into the steam generator, maintains a vacuum at the turbine outlet so that the pressure drop across the turbine, hence the energy extracted from the steam, is maximized. Before being fed into the steam generator, the condensed steam is sometimes preheated in order to minimize thermal shock; the steam generated has other uses besides power generation. In nuclear ships and submarines, the steam is fed through a steam turbine connected to a set of speed reduction gears to a shaft used for propulsion. Direct mechanical action by expansion of the steam can be used for a steam-powered aircraft catapult or similar applications. District heating by the steam is used in some countries and direct heating is applied to internal plant applications. Two things are characteristic for the pressurized water reactor when compared with other reactor types: coolant loop separation from the steam system and pressure inside the primary coolant loop.
In a PWR, there are two separate coolant loops, which are both filled with demineralized/deionized water. A boiling water reactor, by contrast, has only one coolant loop, while more exotic designs such as breeder reactors use substances other than water for coolant and moderator; the pressure in the primary coolant loop is 15–16 megapascals, notably higher than in other nuclear reactors, nearly twice that of a boiling water reactor. As an effect of this, only localized boiling occurs and steam will recondense promptly in the bulk fluid. By contrast, in a boiling water reactor the primary coolant is designed to boil. Light water is used as the primary coolant in a PWR. Water enters through the bottom of the reactor's core at about 548 K and is heated as it flows upwards through the reactor core to a temperature of about 588 K; the water remains liquid despite the high temperature due to the high pressure in the primary coolant loop around 155 bar. In water, the critical point occurs at 22.064 MPa.
Pressure in the primary circuit is maintained by a pressurizer, a separate vessel, conne
Carolinas–Virginia Tube Reactor
Carolinas–Virginia Tube Reactor known as Parr Nuclear Station, was an experimental pressurized tube heavy water nuclear power reactor at Parr, South Carolina in Fairfield County. It was operated by the Carolinas Virginia Nuclear Power Associates. Construction started on January 1, 1960; the CVTR was built to test the concept of a heavy water moderated and cooled pressurized tube reactor for civilian power. It was the first US heavy water power reactor, it was operated by the Carolinas Virginia Nuclear Power Associates, a consortium of the following utilities: Carolina Power & Light Company, Duke Power Company, South Carolina Electric & Gas Company, Virginia Electric and Power Company Design of the CVTR began around 1955. CVTR had a thermal output of about 65 MWth and a gross electrical output of 19 MW. Westinghouse Atomic Power Division was responsible for the design of the nuclear systems while Stone and Webster Engineering designed the remainder of the plant; the reactor consisted of 36 vertical U-tube fuel channels in a moderator tank, 10 feet in diameter and 16 feet tall.
Each leg of the U-tube contained. The reactor used enriched uranium. During power operation, heavy water was circulated by primary pumps through the U-tubes containing the fuel assemblies which heated the water; the heated water flowed through an inverted U-tube steam generator where the heat was transferred to the secondary side light water which turned to steam. The steam flowed to an oil-fired superheater which increased the steam quality before the steam entered the turbine which spun the electrical generator. After passing through the steam generator, the primary loop water was pumped back to the reactor by the primary pumps to repeat the cycle; the primary loop heavy water was pressurized to ensure that the heavy water remained liquid and did not flash to steam at any point in the loop. The U-shaped pressure tubes containing the fuel were thermally isolated from the hot fuel assembly by two circular thermal baffle tubes around the fuel assembly; this allowed the pressure tubes to operate at low temperatures that of the moderator tank, maintained about 155 degrees F and close to atmospheric pressure.
The moderator tank contained heavy water which moderated the fission process during operation of the reactor. The CVTR containment design was a new concept at the time. Designed by Stone and Webster Engineering, the design was focused on not allowing any leakage of radioactive gases or material following an accident; the containment design featured a flat concrete foundation, cylindrical walls, a hemispherical dome, all with an airtight 1/2" or 1/4" steel plate welded liner. The containment was constructed of reinforced concrete. From the basement floor to the interior surface of the top of the dome measured 114’-2”; the vertical walls were 2’-0” thick, the cylindrical structure had an interior diameter of 58’-0”, the dome had a larger interior radius of 29’-4”. The reactor and facilities were located at Parr, SC just to the northeast of the existing Parr Reservoir hydroelectric dam across the Broad River on a high bluff that overlooks the dam powerhouse; the site for the CVTR was approved by the Atomic Energy Commission’s Advisory Committee on Reactor Safeguards in January 1959.
Construction started in 1960. The reactor went critical for the first time on March 30, 1963. CVTR operated from 1963 to 1967, it was shut down after the completion of the planned test program. Staff: Harry Ferguson, General Manager. M. McGough Health Physicist: Lionel Lewis Construction Supervisor: Bill Thomas Engineering Supervisor: Shep Waggoner Following decommissioning of the CVTR, the facility was used for conducting large scale tests to provide experimental information on the response of containment structures to severe events. In the late 1960s, three tests were conducted in which large volumes of steam from the nearby coal-fired power plant was released into the CVTR Containment and the response of the plant measured; the results of these experiments were used for the development and validation of computer model codes. The CVTR has been decommissioned and its license was withdrawn. No fuel remains onsite. By the Fall of 2009, demolition was complete and the site returned to greenfield; the much larger and operational Virgil C.
Summer Nuclear Generating Station was constructed in the 1970s, began operating in 1984 three miles north of the CVTR. Heavy Water Reactors: Status and Projected Development, IAEA Technical Report No. 407 CVTR is described on pp. 52 – 55. Decommissioning Nuclear Power Plants