Generation III reactor

From Wikipedia, the free encyclopedia
https://en.wikipedia.org/wiki/Generation_III_reactor

Model of the Toshiba ABWR, which became the first operational Generation III reactor in 1996

Generation III reactors, or Gen III reactors, are a class of nuclear reactors designed to succeed Generation II reactors, incorporating evolutionary improvements in design. These include improved fuel technology, higher thermal efficiency, significantly enhanced safety systems (including passive nuclear safety), and standardized designs intended to reduce maintenance and capital costs. They are promoted by the Generation IV International Forum (GIF).

The first Generation III reactors to begin operation were Kashiwazaki 6 and 7 advanced boiling water reactors (ABWRs) in 1996 and 1997. From 2012, both have been shut down due to a less permissive political environment in the wake of the Fukushima nuclear accident. Due to the prolonged period of stagnation in the construction of new reactors and the continued (albeit declining) popularity of Generation II/II+ designs in new construction, relatively few third generation reactors have been built.

Overview

The older Gen II reactors comprise the vast majority of current nuclear reactors. Gen III reactors are so-called advanced light-water reactors (LWRs). Gen III+ reactors are labeled as "evolutionary designs". Though the distinction between Gen II and III reactors is arbitrary, few Gen III reactors have reached the commercial stage as of 2022. The Generation IV International Forum calls Gen IV reactors "revolutionary designs". These are concepts for which no concrete prognoses for realization existed at the time. The improvements in reactor technology in third generation reactors are intended to result in a longer operational life (designed for 60 years of operation, extendable to 100+ years of operation prior to complete overhaul and reactor pressure vessel replacement) compared with currently used Generation II reactors (designed for 40 years of operation, extendable to 60+ years of operation prior to complete overhaul and pressure vessel replacement).

The core damage frequencies for these reactors are designed to be lower than for Generation II reactors – 60 core damage events for the European Pressurized Reactor (EPR) and 3 core damage events for the Economic Simplified Boiling Water Reactor (ESBWR) per 100 million reactor-years are significantly lower than the 1,000 core damage events per 100 million reactor-years for BWR/4 Generation II reactors.

The third generation EPR reactor was also designed to use uranium more efficiently than older Generation II reactors, using approximately 17% less per unit of electricity generated than these older reactor technologies. An independent analysis conducted by environmental scientist Barry Brook on the greater efficiency and therefore lower material needs of Gen III reactors, corroborates this finding.

Developments

EPR core catching room designed to catch the corium in case of a meltdown. Some Generation III reactors include a core catcher in their design.

Gen III+ reactor designs are an evolutionary development of Gen III reactors, offering improvements in safety over Gen III reactor designs. Manufacturers began development of Gen III+ systems in the 1990s by building on the operating experience of the American, Japanese, and Western European light-water reactor.

The nuclear industry began to promote a nuclear renaissance suggesting that Gen III+ designs should solve three key problems: safety, cost and buildability. Construction costs of US$1,000/kW were forecast, a level that would make nuclear competitive with gas, and construction times of four years or less were expected. However, these estimates proved over-optimistic.

A notable improvement of Gen III+ systems over second-generation designs is the incorporation in some designs of passive safety features that do not require active controls or operator intervention but instead rely on gravity or natural convection to mitigate the impact of abnormal events.

Kakrapar Atomic Power Station Unit 3 and 4 under construction. India's first Generation III+ reactor

Generation III+ reactors incorporate extra safety features to avoid the kind of disaster suffered at Fukushima in 2011. Generation III+ designs, passive safety, also known as passive cooling, requires no sustained operator action or electronic feedback to shut down the plant safely in the event of an emergency. Many of the Generation III+ nuclear reactors have a core catcher. If the fuel cladding and reactor vessel systems and associated piping become molten, corium will fall into a core catcher which holds the molten material and has the ability to cool it. This, in turn protects the final barrier, the containment building. As an example, Rosatom installed a 200-tonne core catcher in the VVER reactor as the first large piece of equipment in the reactor building of Rooppur 1, describing it as "a unique protection system". In 2017, Rosatom has started commercial operations of the NVNPP-2 Unit 1 VVER-1200 reactor in central Russia, marking the world's first full start-up of a generation III+ reactor.

First reactors

Novovoronezh Nuclear Power Plant II with the first Generation III+ nuclear reactor in the world

The first Generation III reactors were built in Japan, in the form of advanced boiling water reactors. On 5 August 2016, a Generation III+ VVER-1200/392M reactor became operational (first grid connection) at Novovoronezh Nuclear Power Plant II in Russia, which was the first operational Generation III+ reactor.

Several other Generation III+ reactors are under late-stage construction in Europe, China, India, and the United States. The next Generation III+ reactors to come online were an AREVA EPR reactor at the Taishan Nuclear Power Station (first grid connection on 2018-06-29) and a Westinghouse AP1000 reactor at the Sanmen Nuclear Power Station (first grid connection on 2018-06-30) in China.

In the United States, reactor designs are certified by the Nuclear Regulatory Commission (NRC). As of August 2020, the commission has approved seven new designs, and is considering one more design as well as renewal of an expired certification.

Response and criticism

Proponents of nuclear power and some who have historically been critical have acknowledged that third generation reactors as a whole are safer than older reactors.

Edwin Lyman, a senior staff scientist at the Union of Concerned Scientists, has challenged specific cost-saving design choices made for two Generation III reactors, both the AP1000 and ESBWR. Lyman, John Ma (a senior structural engineer at the NRC), and Arnold Gundersen (an anti-nuclear consultant) are concerned about what they perceive as weaknesses in the steel containment vessel and the concrete shield building around the AP1000 in that its containment vessel does not have sufficient safety margins in the event of a direct airplane strike. Other engineers do not agree with these concerns, and claim the containment building is more than sufficient in safety margins and factors of safety.

The Union of Concerned Scientists in 2008 referred to the EPR as the only new reactor design under consideration in the United States that "...appears to have the potential to be significantly safer and more secure against attack than today's reactors."

There have also been issues in fabricating the precision parts necessary to maintain safe operation of these reactors, with cost overruns, broken parts, and extremely fine steel tolerances causing issues with new reactors under construction in France at the Flamanville Nuclear Power Plant.

Lists of Generation III reactors

Generation III reactors currently operational or under construction

Generation III designs not adopted or built yet

Lists of Generation III+ reactors

Generation III+ reactors currently operational or under construction

Generation III+ designs not adopted or built yet

Developer(s)	Reactor name(s)	Type	MWe (net)	MWe (gross)	MWth	Notes
General Electric, Toshiba, Hitachi	ABWR; US-ABWR	BWR	1350	1420	3926	In operation at Kashiwazaki since 1996. NRC certified in 1997.
KEPCO	APR-1400	PWR	1383	1455	3983	In operation at Kori since Jan 2016.
CGNPG	ACPR-1000		1061	1119	2905	Improved version of the CPR-1000. The first reactor came online in 2018 at Yangjiang-5.
CGNPG, CNNC	Hualong One (HPR-1000)		1090	1170	3050	In part a merger of the Chinese ACPR-1000 and ACP-1000 designs, but ultimately an incrementally developed improvement on the prior CNP-1000 and CP-1000 designs. It was initially intended to be named the "ACC-1000", but was ultimately named as the "Hualong One" or "HPR-1000". Fangchenggang Units 3–6 will be the first to utilize the HPR-1000 design, with Units 3 & 4 currently under construction as of 2017.
OKBM Afrikantov	VVER-1000/428		990	1060	3000	First version of the AES-91 design, designed and used for Tianwan Units 1 & 2, which came online in 2007.
	VVER-1000/428M		1050	1126	3000	Another version of the AES-91 design, also designed and used for Tianwan (this time for Units 3 & 4, which came online in 2017 and 2018, respectively).
	VVER-1000/412		917	1000	3000	First constructed AES-92 design, used for the Kudankulam.
Developer(s)	Reactor name(s)	Type	MWe (net)	MWe (gross)	MWth	Notes
General Electric, Hitachi	ABWR-II	BWR	1638	1717	4960	Improved version of the ABWR. Uncertain development status.
Mitsubishi	APWR; US-APWR; EU-APWR; APWR+	PWR	1600	1700	4451	Two units planned at Tsuruga cancelled in 2011. US NRC licensing for two units planned at Comanche Peak was suspended in 2013. The original APWR and the updated US-APWR/EU-APWR (also known as the APWR+) differ significantly in their design characteristics, with the APWR+ having higher efficiency and electrical output.
Westinghouse	AP600		600	619	?	NRC certified in 1999. Evolved into the larger AP1000 design.
Combustion Engineering	System 80+		1350	1400	?	NRC certified in 1997. Provided a basis for the Korean APR-1400.
OKBM Afrikantov	VVER-1000/466(B)		1011	1060	3000	This was the first AES-92 design to be developed, originally intended to be built at the proposed Belene Nuclear Power Plant, but construction was later halted.
Candu Energy Inc.	EC6	PHWR	?	750	2084	The EC6 (Enhanced CANDU 6) is an evolutionary upgrade of previous CANDU designs. Like other CANDU designs, it is capable of using unenriched natural uranium as fuel.
	AFCR		?	740	2084	The Advanced Fuel CANDU Reactor is a modified EC6 design that has been optimized for extreme fuel flexibility with the ability to handle numerous potential reprocessed fuel blends and even thorium. It is currently undergoing late-stage development as part of a joint venture between SNC-Lavalin, CNNC, and Shanghai Electric.
Various (see MKER Article.)	MKER	BWR	1000	?	2085	A Development of the RBMK nuclear power reactor. Fixes all of the RBMK reactor's design errors and flaws and adds a full containment building and Passive nuclear safety features such as a passive core cooling system. The physical prototype of the MKER-1000 is the 5th unit of the Kursk Nuclear Power Plant. The construction of Kursk 5 was cancelled in 2012 and a VVER-TOI whose construction is ongoing since 2018 is being built instead as of 2018.
Developer(s)	Reactor name(s)	Type	MWe (net)	MWe (gross)	MWth	First grid connection	Notes
Westinghouse, Toshiba	AP1000	PWR	1117	1250	3400	2018-06-30 Sanmen	NRC certified Dec 2005.
SNPTC, Westinghouse	CAP1400		1400	1500	4058		The first Chinese co-developed and upsized "native" version/derivative of the AP1000. Westinghouse's co-development agreement gives China the IP rights for all co-developed plants >1350 MWe. First two units currently under construction at Shidao Bay. The CAP1400 is planned to be followed by a CAP1700 and/or a CAP2100 design if the cooling systems can be scaled up by far enough.
Areva	EPR		1660	1750	4590	2018-06-29 Taishan
OKB Gidropress	VVER-1200/392M		1114	1180	3200	2016-08-05 Novovoronezh II	The VVER-1200 series is also known as the AES-2006/MIR-1200 design. This particular model was the original reference model used for the VVER-TOI project.
	VVER-1200/491		1085	1199	3200	2018-03-09 Leningrad II
	VVER-1200/509		1114	1200	3200		Under construction in Akkuyu NPP, as Akkuyu 1 & 2. Grid connections due 2023 & 2024.
	VVER-1200/523		1080	1200	3200		2.4 GWe Rooppur Nuclear Power Plant of Bangladesh is under construction.The two units of VVER- 1200/523 generating 2.4 GWe are planned to be operational in 2023 and 2024.
	VVER-1200/513		?	1200	3200		Standardized version of the VVER-1200 based in part on the VVER-1300/510 design (which is the current reference design for the VVER-TOI project). First unit expected to be completed by 2022 at Akkuyu, as Akkuyu 3.
	VVER-1300/510		1115	1255	3300		The VVER-1300 design is also known as the AES-2010 design, and is sometimes mistakenly designated as the VVER-TOI design. The VVER-1300/510 is based on the VVER-1200/392M that was originally used as the reference design for the VVER-TOI project, although the VVER-1300/510 now serves that role (which has led to confusion between the VVER-TOI plant design and the VVER-1300/510 reactor design). Multiple units are currently planned for construction at several Russian nuclear plants. First units under construction at Kursk Nuclear Power Plant.
BARC	IPHWR-700	PHWR	630	700	2166	2021	Successor of indigenous 540MWe PHWR with increased output and additional safety features. Under construction and due to come online in 2020. Unit 3 at Kakrapar Atomic Power Station achieved first criticality on 22 July 2020. The Unit 3 was connected to the grid on 10 January 2021.
Developer(s)	Reactor name(s)	Type	MWe (net)	MWe (gross)	MWth	Notes
Toshiba	EU-ABWR	BWR	?	1600	4300	Updated version of the ABWR designed to meet EU guidelines, increase reactor output, and improve design generation to III+.
Areva	Kerena		1250	1290	3370	Previously known as the SWR-1000. Based on German BWR designs, mainly that of Gundremmingen units B/C. Co-developed by Areva and E.ON.
General Electric, Hitachi	ESBWR		1520	1600	4500	Based on the unreleased SBWR design which in turn was based on the ABWR. Being considered for North Anna-3. Eschews the use of recirculation pumps entirely in favor of a design completely reliant on natural circulation (which is very unusual for a boiling water reactor design).
KEPCO	APR+	PWR	1505	1560	4290	APR-1400 successor with increased output and additional safety features.
Areva, Mitsubishi	ATMEA1		1150	?	3150	Proposed Sinop plant did not proceed
OKB Gidropress	VVER-600/498		?	600	1600	Essentially a scaled-down VVER-1200. Commercial deployment planned by 2030 at Kola.
Candu Energy Inc.	ACR-1000	PHWR	1085	1165	3200	The Advanced CANDU Reactor is a hybrid CANDU design that retains the heavy water moderator but replaces the heavy water coolant with conventional light water coolant, significantly reducing heavy water costs compared to traditional CANDU designs but losing the characteristic CANDU capability of using unenriched natural uranium as fuel.
BARC	IPWR-900	PWR	900	?	2700	India's first light water reactor, a Gen 3+ design based on the CLWR-B1 reactor of Arihant-class submarine.

Supercritical water reactor

From Wikipedia, the free encyclopedia
https://en.wikipedia.org/wiki/Supercritical_water_reactor

Supercritical water reactor scheme.

The supercritical water reactor (SCWR) is a concept Generation IV reactor, designed as a light water reactor (LWR) that operates at supercritical pressure (i.e. greater than 22.1 megapascals [3,210 psi]). The term critical in this context refers to the critical point of water, and should not be confused with the concept of criticality of the nuclear reactor.

The water heated in the reactor core becomes a supercritical fluid above the critical temperature of 374 °C (705 °F), transitioning from a fluid more resembling liquid water to a fluid more resembling saturated steam (which can be used in a steam turbine), without going through the distinct phase transition of boiling.

In contrast, the well-established pressurized water reactors (PWR) have a primary cooling loop of liquid water at a subcritical pressure, transporting heat from the reactor core to a secondary cooling loop, where the steam for driving the turbines is produced in a boiler (called the steam generator). Boiling water reactors (BWR) operate at even lower pressures, with the boiling process to generate the steam happening in the reactor core.

The supercritical steam generator is a proven technology.

The development of SCWR systems is considered a promising advancement for nuclear power plants because of its high thermal efficiency (~45 % vs. ~33 % for current LWRs) and simpler design. As of 2012 the concept was being investigated by 32 organizations in 13 countries.

History

The super-heated steam cooled reactors operating at subcritical-pressure were experimented with in both Soviet Union and in the United States as early as the 1950s and 1960s such as Beloyarsk Nuclear Power Station, Pathfinder and Bonus of GE's Operation Sunrise program. These are not SCWRs. SCWRs were developed from the 1990s onwards. Both a LWR-type SCWR with a reactor pressure vessel and a CANDU-type SCWR with pressure tubes are being developed.

A 2010 book includes conceptual design and analysis methods such as core design, plant system, plant dynamics and control, plant startup and stability, safety, fast reactor design etc.

A 2013 document saw the completion of a prototypical fueled loop test in 2015. A Fuel Qualification Test was completed in 2014.

A 2014 book saw reactor conceptual design of a thermal spectrum reactor (Super LWR) and a fast reactor (Super FR) and experimental results of thermal hydraulics, materials and material-coolant interactions.

Design

Moderator-coolant

The SCWR operates at supercritical pressure. The reactor outlet coolant is supercritical water. Light water is used as a neutron moderator and coolant. Above the critical point, steam and liquid become the same density and are indistinguishable, eliminating the need for pressurizers and steam generators (PWR), or jet/recirculation pumps, steam separators and dryers (BWR). Also, by avoiding boiling, SCWR does not generate chaotic voids (bubbles) with less density and moderating effect. In a LWR this can affect heat transfer and water flow, and the feedback can make the reactor power harder to predict and control. Neutronic and thermal hydraulic coupled calculation is needed to predict the power distribution. SCWR's simplification should reduce construction costs and improve reliability and safety.

A LWR type SCWR adopts water rods with thermal insulation and a CANDU type SCWR keeps water moderator in a Calandria tank. A fast reactor core of the LWR type SCWR adopts tight fuel rod lattice as a high conversion LWR. The fast neutron spectrum SCWR has advantages of a higher power density, but needs plutonium and uranium mixed oxides fuel which will be available from reprocessing.　

Control

SCWRs would likely have control rods inserted through the top, as is done in PWRs.

Material

The temperature inside an SCWR is higher than those in LWRs. Although supercritical fossil fuel plants have much experience in the materials, it does not include the combination of high temperature environment and intense neutron radiation. SCWRs need core materials (especially fuel cladding) to resist the environment. R&D focuses on:

• The chemistry of supercritical water under radiation (preventing stress corrosion cracking, and maintaining corrosion resistance under neutron radiation and high temperatures)
• Dimensional and microstructural stability (preventing embrittlement, retaining strength and creep resistance also under radiation and high temperatures)
• Materials that both resist the high temperature conditions and do not absorb too many neutrons, which affects fuel economy

In the once-through coolant cycles, such as SCWRs and supercritical fossil fired power plants, the entire reactor coolant is processed at low temperature after condensation. It is an advantage in managing water chemistry and stress corrosion cracking of structural materials. It is not possible in LWRs due to the recirculation of hot reactor coolant. Materials and water chemistry R&D should be done with the once-through characteristics in mind.

Advantages

• Supercritical water has excellent heat transfer properties allowing a high-power density, a small core, and a small containment structure.
• The use of a supercritical Rankine cycle with its typically higher temperatures improves efficiency (would be ~45 % versus ~33 % of current PWR/BWRs).
• This higher efficiency would lead to better fuel economy and a lighter fuel load, lessening residual (decay) heat.
• SCWR is typically designed as a direct cycle, whereby steam or hot supercritical water from the core is used directly in a steam turbine. This makes the design simple. As a BWR is simpler than a PWR, a SCWR is a lot simpler and more compact than a less-efficient BWR having the same electrical output. There are no steam separators, steam dryers, internal recirculation pumps, or recirculation flow inside the pressure vessel. The design is a once-through, direct-cycle, the simplest type of cycle possible. The stored thermal and radiologic energy in the smaller core and its (primary) cooling circuit would also be less than that of either a BWR's or a PWR's.
• Water is liquid at room temperature, cheap, non-toxic and transparent, simplifying inspection and repair (compared to liquid metal cooled reactors).
• A fast SCWR could be a breeder reactor, like the proposed Clean and Environmentally Safe Advanced Reactor and could burn the long-lived actinide isotopes.
• A heavy-water SCWR could breed fuel from thorium (4x more abundant than uranium). Similar to a CANDU it could also use unenriched natural uranium if enough moderation is provided
• Process heat can be delivered at higher temperatures than other water-cooled reactors allow

Disadvantages

• Lower water inventory (due to compact primary loop) means less heat capacity to buffer transients and accidents (e.g., loss of feedwater flow or large break loss-of-coolant accident) resulting in accident and transient temperatures that are too high for conventional metallic cladding.

However, it is not too high for stainless steel cladding. Safety analysis of LWR type SCWR showed that safety criteria are met with margins at accidents and abnormal transients including total loss of flow and loss of coolant accident. No double ended break occurs because of the once-through coolant cycle. Core is cooled by the induced flow at the loss of coolant accident. The water inventory in the top dome of the reactor vessel serves as an in-vessel accumulator. The SCWR safety principle is not to maintain coolant inventory, but to maintain core coolant flow rate. It is easier to monitor than water level at accidents. There was an error in the water level signal in the Three Mile Island accident and the operators shut down the ECCS.

• Higher pressure combined with higher temperature and also a higher temperature rises across the core (compared to PWR/BWRs) result in increased mechanical and thermal stresses on vessel materials that are difficult to solve.

However, a LWR type design, reactor pressure vessel inner wall is cooled by the inlet coolant as PWR. Outlet coolant nozzles are equipped with thermal sleeves. A pressure-tube design, where the core is divided up into smaller tubes for each fuel channel, has potentially fewer issues here, as smaller diameter tubing can be much thinner than massive single pressure vessels, and the tube can be insulated on the inside with inert ceramic insulation so it can operate at low (calandria water) temperature.

• The coolant greatly reduces its density at the end of the core, resulting in a need to place extra moderator there.

However, a LWR type SCWR design adopts water rods in the fuel assemblies as BWRs. The coolant density in water rods is kept high with thin thermal insulation, not fully insulated. Most designs of CANDU type SCWR use an internal calandria where part of the feedwater flow is guided through top tubes through the core, that provide the added moderation (feedwater) in that region. This has the added advantage of being able to cool the entire vessel wall with feedwater, but results in a complex and materially demanding (high temperature, high temperature differences, high radiation) internal calandria and plena arrangement. A pressure-tube design has the characteristics as most of the moderator is in the calandria at low temperature and pressure, reducing the coolant density effect on moderation, and the actual pressure tube can be kept cool by the calandria water.

• Extensive material development and research on supercritical water chemistry under radiation is needed.

However, the entire SCWR coolant is cleaned after condensation. This is an advantage in managing water chemistry and Stress corrosion cracking of structural materials. It is not possible in LWRs where hot coolant circulates.

• Special start-up procedures needed to avoid instability before the water reaches supercritical conditions.

However, Instability is managed by power to coolant flow rate ratio as a BWR. The coolant density change is smaller in SCWRs than BWRs.

• A fast SCWR needs a relatively complex reactor core to have a negative void coefficient.

However, single coolant flow pass core is feasible.

• As with all alternatives to currently widespread designs (mostly subcritical water cooled, water moderated thermal reactors of some kind) there will be fewer suppliers of technology and parts and less expertise at least initially than for decades old proven technology or its evolutionary improvements such as generation III+ reactors.

However, LWRs were developed in the 1950s based on the subcritical fossil fired power technologies. The success of LWRs is based on that experience. Supercritical fossil fired power plants were developed after 1950s. Components such as valves, piping, turbines, feedwater pumps and heaters for operation at turbine throttle pressure up to 30 MPa (4,400 psi) and temperature up to 630 °C (903 K; 1,166 °F) are present for commercial applications. SCWRs are natural evolution of LWRs. The competitiveness of LWRs in the electricity market is being challenged in the US due to Shale gas from historical summaries of U.S. Energy Information Administration’s (EIA’s) Levelized cost of electricity (LCOE) projections (2010-2020) in Cost of electricity by source. LWRs are the dominant design with the largest share of nuclear power generation and are the current offering for new construction in the world. Innovation dynamics show that innovation does not come from companies with the largest market share. Comparing SCWRs and LWRs is not relevant in terms of innovation dynamics. If Small modular reactor (SMR) is competitive, a SMR version of SCWRs will increase its advantage.

• The chemical shim might behave drastically different as the solution properties of supercritical water are vastly different from those of liquid water. Currently most pressurized water reactors employ boric acid to control reactivity early in burnup.

However, chemical shim cannot be used in SCWRs as well as BWRs, due to the positive coolant void coefficient. SCWRs use borated water as the secondary shut-down similar to BWRs.

• Depending on design online refuelling may be impossible. While CANDUs are capable of online refuelling, other water moderated reactors are not.

However, the Capacity factor of LWRs is already high in USA, over 90%. Pressure vessel type SCWRs do not require online refuelling.

Pressurized water reactor

From Wikipedia, the free encyclopedia
https://en.wikipedia.org/wiki/Pressurized_water_reactor

 

Nuclear Regulatory Commission image of pressurized water reactor vessel heads

A pressurized water reactor (PWR) is a type of light-water nuclear reactor. PWRs constitute the large majority of the world's nuclear power plants (with notable exceptions being the UK, Japan and Canada). In a PWR, the primary coolant (water) is pumped under high pressure to the reactor core where it is heated by the energy released by the fission of atoms. The heated, high pressure water then flows to a steam generator, where it transfers its thermal energy to lower pressure water of a secondary system where steam is generated. The steam then drives turbines, which spin an electric generator. In contrast to a boiling water reactor (BWR), pressure in the primary coolant loop prevents the water from boiling within the reactor. All light-water reactors use ordinary water as both coolant and neutron moderator. Most use anywhere from two to four vertically mounted steam generators; VVER reactors use horizontal steam generators.

PWRs were originally 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 currently operating in the United States are considered Generation II reactors. Russia's VVER reactors are similar to US PWRs, but the VVER-1200 is not considered Generation II (see below). France operates many PWRs to generate the bulk of its electricity.

History

Rancho Seco PWR reactor hall and cooling tower (being decommissioned, 2004)

Several hundred PWRs are used for marine propulsion in aircraft carriers, nuclear submarines and ice breakers. In the US, they were originally designed at the Oak Ridge National Laboratory for use as a nuclear submarine power plant with a fully operational submarine power plant located at the Idaho National Laboratory. Follow-on work was conducted by Westinghouse Bettis Atomic Power Laboratory. The first purely commercial nuclear power plant at Shippingport Atomic Power Station was originally designed as a pressurized water reactor (although the first power plant connected to the grid was at Obninsk, USSR), 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 initially operated two pressurized water reactor plants, TMI-1 and TMI-2. The partial meltdown of TMI-2 in 1979 essentially ended the growth in new construction of nuclear power plants in the United States for two decades. Watts Bar unit 2 (a Westinghouse 4-loop PWR) came online in 2016, becoming the first new nuclear reactor in the United States since 1996.

The pressurized water reactor has several new Generation III reactor evolutionary designs: the AP1000, VVER-1200, ACPR1000+, APR1400, Hualong One, IPWR-900 and EPR. The first AP1000 and EPR reactors were connected to the power grid in China in 2018. In 2020, NuScale Power became the first U.S. company to receive regulatory approval from the Nuclear Regulatory Commission for a small modular reactor with a modified PWR design. Also in 2020, the Energy Impact Center introduced the OPEN100 project, which published open-source blueprints for the construction of a 100 MW_electric nuclear power plant with a PWR design.

Design

Pictorial explanation of power transfer in a pressurized water reactor. Primary coolant is in orange and the secondary coolant (steam and later feedwater) is in blue.

Primary coolant system showing reactor pressure vessel (red), steam generators (purple), Pressurizer (blue), and pumps (green) in the three coolant loop Hualong One design

Nuclear fuel in the reactor pressure vessel is engaged in a controlled 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 several thousand small tubes. Heat is transferred through the walls of these tubes to the lower pressure secondary coolant located on the shell side of the exchanger where the secondary coolant evaporates to pressurized steam. This transfer of heat is accomplished without mixing the two fluids to prevent the secondary coolant from becoming radioactive. Some common steam generator arrangements are u-tubes or 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 (water-steam mixture) is cooled down and condensed in a condenser. The condenser converts the steam to a liquid so that it can be pumped back into the steam generator, and maintains a vacuum at the turbine outlet so that the pressure drop across the turbine, and hence the energy extracted from the steam, is maximized. Before being fed into the steam generator, the condensed steam (referred to as feedwater) 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 (PWR) 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 (primary and secondary), 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 (e.g. sodium in its liquid state as coolant or graphite as a moderator). The pressure in the primary coolant loop is typically 15–16 megapascals (150–160 bar), which is notably higher than in other nuclear reactors, and nearly twice that of a boiling water reactor (BWR). 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.

Reactor

PWR reactor pressure vessel

Coolant

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 (275 °C; 527 °F) and is heated as it flows upwards through the reactor core to a temperature of about 588 K (315 °C; 599 °F). The water remains liquid despite the high temperature due to the high pressure in the primary coolant loop, usually around 155 bar (15.5 MPa 153 atm, 2,250 psi). The water in a PWR cannot exceed a temperature of 647 K (374 °C; 705 °F) or a pressure of 22.064 MPa (3200 psi or 218 atm), because those are the critical point of water. Supercritical water reactors are (as of 2022) only a proposed concept in which the coolant would never leave the supercritical state. However, as this requires even higher pressures than a PWR and can cause issues of corrosion, so far no such reactor has been built.

Pressurizer

Main article: Pressurizer

Pressure in the primary circuit is maintained by a pressurizer, a separate vessel that is connected to the primary circuit and partially filled with water which is heated to the saturation temperature (boiling point) for the desired pressure by submerged electrical heaters. To achieve a pressure of 155 bars (15.5 MPa), the pressurizer temperature is maintained at 345 °C (653 °F), which gives a subcooling margin (the difference between the pressurizer temperature and the highest temperature in the reactor core) of 30 °C (54 °F). As 345 °C is the boiling point of water at 155 bar, the liquid water is at the edge of a phase change. Thermal transients in the reactor coolant system result in large swings in pressurizer liquid/steam volume, and total pressurizer volume is designed around absorbing these transients without uncovering the heaters or emptying the pressurizer. Pressure transients in the primary coolant system manifest as temperature transients in the pressurizer and are controlled through the use of automatic heaters and water spray, which raise and lower pressurizer temperature, respectively.

Pumps

The coolant is pumped around the primary circuit by powerful pumps. These pumps have a rate of ~100,000 gallons of coolant per minute. After picking up heat as it passes through the reactor core, the primary coolant transfers heat in a steam generator to water in a lower pressure secondary circuit, evaporating the secondary coolant to saturated steam — in most designs 6.2 MPa (60 atm, 900 psia), 275 °C (530 °F) — for use in the steam turbine. The cooled primary coolant is then returned to the reactor vessel to be heated again.

Moderator

Main article: Passive nuclear safety

Pressurized water reactors, like all thermal reactor designs, require the fast fission neutrons to be slowed (a process called moderation or thermalizing) in order to interact with the nuclear fuel and sustain the chain reaction. In PWRs the coolant water is used as a moderator by letting the neutrons undergo multiple collisions with light hydrogen atoms in the water, losing speed in the process. This "moderating" of neutrons will happen more often when the water is more dense (more collisions will occur). The use of water as a moderator is an important safety feature of PWRs, as an increase in temperature may cause the water to expand, giving greater 'gaps' between the water molecules and reducing the probability of thermalization — thereby reducing the extent to which neutrons are slowed and hence reducing the reactivity in the reactor. Therefore, if reactivity increases beyond normal, the reduced moderation of neutrons will cause the chain reaction to slow down, producing less heat. This property, known as the negative temperature coefficient of reactivity, makes PWR reactors very stable. This process is referred to as 'Self-Regulating', i.e. the hotter the coolant becomes, the less reactive the plant becomes, shutting itself down slightly to compensate and vice versa. Thus the plant controls itself around a given temperature set by the position of the control rods.

In contrast, the Soviet RBMK reactor design used at Chernobyl, which uses graphite instead of water as the moderator and uses boiling water as the coolant, has a large positive thermal coefficient of reactivity. This means reactivity and heat generation increases when coolant and fuel temperatures increase, which makes the RBMK design less stable than pressurized water reactors at high operating temperature. In addition to its property of slowing down neutrons when serving as a moderator, water also has a property of absorbing neutrons, albeit to a lesser degree. When the coolant water temperature increases, the boiling increases, which creates voids. Thus there is less water to absorb thermal neutrons that have already been slowed by the graphite moderator, causing an increase in reactivity. This property is called the void coefficient of reactivity, and in an RBMK reactor like Chernobyl, the void coefficient is positive, and fairly large, making it very hard to regulate when the reaction begins to run away. The RBMK reactors also have a flawed control rods design in which during rapid scrams, the graphite reaction enhancement tips of the rods would displace water at the bottom of the reactor and locally increase reactivity there. This is called the "positive scram effect" that is unique to the flawed RBMK control rods design. These design flaws, in addition to operator errors that pushed the reactor to its limits, are generally seen as the causes of the Chernobyl disaster. 

The Canadian CANDU heavy water reactor design have a slight positive void coefficient, these reactors mitigate this issues with a number of built-in advanced passive safety systems not found in the Soviet RBMK design. No criticality could occur in a CANDU reactor or any other heavy water reactor when ordinary light water is supplied to the reactor as an emergency coolant. Depending on burnup, boric acid or another neutron poison will have to be added to emergency coolant to avoid a criticality accident.

PWRs are designed to be maintained in an undermoderated state, meaning that there is room for increased water volume or density to further increase moderation, because if moderation were near saturation, then a reduction in density of the moderator/coolant could reduce neutron absorption significantly while reducing moderation only slightly, making the void coefficient positive. Also, light water is actually a somewhat stronger moderator of neutrons than heavy water, though heavy water's neutron absorption is much lower. Because of these two facts, light water reactors have a relatively small moderator volume and therefore have compact cores. One next generation design, the supercritical water reactor, is even less moderated. A less moderated neutron energy spectrum does worsen the capture/fission ratio for ²³⁵U and especially ²³⁹Pu, meaning that more fissile nuclei fail to fission on neutron absorption and instead capture the neutron to become a heavier nonfissile isotope, wasting one or more neutrons and increasing accumulation of heavy transuranic actinides, some of which have long half-lives.

Fuel

Main article: Nuclear fuel

PWR fuel bundle This fuel bundle is from a pressurized water reactor of the nuclear passenger and cargo ship NS Savannah. Designed and built by Babcock & Wilcox.

After enrichment, the uranium dioxide (UO
₂) powder is fired in a high-temperature, sintering furnace to create hard, ceramic pellets of enriched uranium dioxide. The cylindrical pellets are then clad in a corrosion-resistant zirconium metal alloy Zircaloy which are backfilled with helium to aid heat conduction and detect leakages. Zircaloy is chosen because of its mechanical properties and its low absorption cross section. The finished fuel rods are grouped in fuel assemblies, called fuel bundles, that are then used to build the core of the reactor. A typical PWR has fuel assemblies of 200 to 300 rods each, and a large reactor would have about 150–250 such assemblies with 80–100 tons of uranium in all. Generally, the fuel bundles consist of fuel rods bundled 14 × 14 to 17 × 17. A PWR produces on the order of 900 to 1,600 MW_e. PWR fuel bundles are about 4 meters in length.

Refuelings for most commercial PWRs is on an 18–24 month cycle. Approximately one third of the core is replaced each refueling, though some more modern refueling schemes may reduce refuel time to a few days and allow refueling to occur on a shorter periodicity.

Control

In PWRs reactor power can be viewed as following steam (turbine) demand due to the reactivity feedback of the temperature change caused by increased or decreased steam flow. (See: Negative temperature coefficient.) Boron and cadmium control rods are used to maintain primary system temperature at the desired point. In order to decrease power, the operator throttles shut turbine inlet valves. This would result in less steam being drawn from the steam generators. This results in the primary loop increasing in temperature. The higher temperature causes the density of the primary reactor coolant water to decrease, allowing higher neutron speeds, thus less fission and decreased power output. This decrease of power will eventually result in primary system temperature returning to its previous steady-state value. The operator can control the steady state operating temperature by addition of boric acid and/or movement of control rods.

Reactivity adjustment to maintain 100% power as the fuel is burned up in most commercial PWRs is normally achieved by varying the concentration of boric acid dissolved in the primary reactor coolant. Boron readily absorbs neutrons and increasing or decreasing its concentration in the reactor coolant will therefore affect the neutron activity correspondingly. An entire control system involving high pressure pumps (usually called the charging and letdown system) is required to remove water from the high pressure primary loop and re-inject the water back in with differing concentrations of boric acid. The reactor control rods, inserted through the reactor vessel head directly into the fuel bundles, are moved for the following reasons: to start up the reactor, to shut down the primary nuclear reactions in the reactor, to accommodate short term transients, such as changes to load on the turbine,

The control rods can also be used to compensate for nuclear poison inventory and to compensate for nuclear fuel depletion. However, these effects are more usually accommodated by altering the primary coolant boric acid concentration.

In contrast, BWRs have no boron in the reactor coolant and control the reactor power by adjusting the reactor coolant flow rate.

Advantages

PWR reactors are very stable due to their tendency to produce less power as temperatures increase; this makes the reactor easier to operate from a stability standpoint.

PWR turbine cycle loop is separate from the primary loop, so the water in the secondary loop is not contaminated by radioactive materials.

PWRs can passively scram the reactor in case offsite power is lost to immediately stop the primary nuclear reaction. The control rods are held by electromagnets and fall by gravity when current is lost; full insertion safely shuts down the primary nuclear reaction.

PWR technology is favoured by nations seeking to develop a nuclear navy; the compact reactors fit well in nuclear submarines and other nuclear ships.

PWRs are the most deployed type of reactor globally, allowing for a wide range of suppliers of new plants and parts for existing plants. Due to long experience with their operation they are the closest thing to mature technology that exists in nuclear energy.

PWRs - depending on type - can be fueled with MOX-fuel and/or the Russian Remix Fuel (which has a lower ²³⁹
Pu and a higher ²³⁵
U content than "regular" U/Pu MOX-fuel) allowing for a (partially) closed nuclear fuel cycle.

Water is a nontoxic, transparent, chemically unreactive (by comparison with e.g. NaK) coolant that is liquid at room temperature which makes visual inspection and maintenance easier. It is also easy and cheap to obtain unlike heavy water or even nuclear graphite.

Compared to reactors operating on natural uranium, PWRs can achieve a relatively high burnup. A typical PWR will exchange a quarter to a third of its fuel load every 18-24 months and have maintenance and inspection, that requires the reactor to be shut down, scheduled for this window. While more uranium ore is consumed per unit of electricity produced than in a natural uranium fueled reactor, the amount of spent fuel is less with the balance being depleted uranium whose radiological danger is lower than that of natural uranium.

Disadvantages

The coolant water must be highly pressurized to remain liquid at high temperatures. This requires high strength piping and a heavy pressure vessel and hence increases construction costs. The higher pressure can increase the consequences of a loss-of-coolant accident. The reactor pressure vessel is manufactured from ductile steel but, as the plant is operated, neutron flux from the reactor causes this steel to become less ductile. Eventually the ductility of the steel will reach limits determined by the applicable boiler and pressure vessel standards, and the pressure vessel must be repaired or replaced. This might not be practical or economic, and so determines the life of the plant.

Additional high pressure components such as reactor coolant pumps, pressurizer, and steam generators are also needed. This also increases the capital cost and complexity of a PWR power plant.

The high temperature water coolant with boric acid dissolved in it is corrosive to carbon steel (but not stainless steel); this can cause radioactive corrosion products to circulate in the primary coolant loop. This not only limits the lifetime of the reactor, but the systems that filter out the corrosion products and adjust the boric acid concentration add significantly to the overall cost of the reactor and to radiation exposure. In one instance, this has resulted in severe corrosion to control rod drive mechanisms when the boric acid solution leaked through the seal between the mechanism itself and the primary system.

Due to the requirement to load a pressurized water reactor's primary coolant loop with boron, undesirable radioactive secondary tritium production in the water is over 25 times greater than in boiling water reactors of similar power, owing to the latter's absence of the neutron moderating element in its coolant loop. The tritium is created by the absorption of a fast neutron in the nucleus of a boron-10 atom which subsequently splits into a lithium-7 and tritium atom. Pressurized water reactors annually emit several hundred curies of tritium to the environment as part of normal operation.

Natural uranium is only 0.7% uranium-235, the isotope necessary for thermal reactors. This makes it necessary to enrich the uranium fuel, which significantly increases the costs of fuel production. Compared to reactors operating on natural uranium, less energy is generated per unit of uranium ore even though a higher burnup can be achieved. Nuclear reprocessing can "stretch" the fuel supply of both natural uranium and enriched uranium reactors but is virtually only practiced for light water reactors operating with lightly enriched fuel as spent fuel from e.g. CANDU reactors is very low in fissile material.

Because water acts as a neutron moderator, it is not possible to build a fast-neutron reactor with a PWR design. A reduced moderation water reactor may however achieve a breeding ratio greater than unity, though this reactor design has disadvantages of its own.

Spent fuel from a PWR usually has a higher content of fissile material than natural uranium. Without nuclear reprocessing, this fissile material cannot be used as fuel in a PWR. It can, however, be used in a CANDU with only minimal reprocessing in a process called "DUPIC" - Direct Use of spent PWR fuel in CANDU.

Thermal efficiency, while better than for boiling water reactors, cannot achieve the values of reactors with higher operating temperatures such as those cooled with high temperature gases, liquid metals or molten salts. Similarly process heat drawn from a PWR is not suitable for most industrial applications as those require temperatures in excess of 400 °C (752 °F).

Radiolysis and certain accident scenarios which involve interactions between hot steam and zircalloy cladding can produce hydrogen from the cooling water leading to hydrogen explosions as a potential accident scenario. During the Fukushima nuclear accident a hydrogen explosion damaging the containment building was a major concern. Some reactors contain catalytic recombiners which let the hydrogen react with ambient oxygen in a non-explosive fashion.

Neurotoxicity

From Wikipedia, the free encyclopedia
https://en.wikipedia.org/wiki/Neurotoxicity

Neurotoxicity is a form of toxicity in which a biological, chemical, or physical agent produces an adverse effect on the structure or function of the central and/or peripheral nervous system. It occurs when exposure to a substance – specifically, a neurotoxin or neurotoxicant– alters the normal activity of the nervous system in such a way as to cause permanent or reversible damage to nervous tissue. This can eventually disrupt or even kill neurons, which are cells that transmit and process signals in the brain and other parts of the nervous system. Neurotoxicity can result from organ transplants, radiation treatment, certain drug therapies, recreational drug use, exposure to heavy metals, bites from certain species of venomous snakes, pesticides, certain industrial cleaning solvents, fuels and certain naturally occurring substances. Symptoms may appear immediately after exposure or be delayed. They may include limb weakness or numbness, loss of memory, vision, and/or intellect, uncontrollable obsessive and/or compulsive behaviors, delusions, headache, cognitive and behavioral problems and sexual dysfunction. Chronic mold exposure in homes can lead to neurotoxicity which may not appear for months to years of exposure. All symptoms listed above are consistent with mold mycotoxin accumulation.

The term neurotoxicity implies the involvement of a neurotoxin; however, the term neurotoxic may be used more loosely to describe states that are known to cause physical brain damage, but where no specific neurotoxin has been identified.

The presence of neurocognitive deficits alone is not usually considered sufficient evidence of neurotoxicity, as many substances may impair neurocognitive performance without resulting in the death of neurons. This may be due to the direct action of the substance, with the impairment and neurocognitive deficits being temporary, and resolving when the substance is eliminated from the body. In some cases the level or exposure-time may be critical, with some substances only becoming neurotoxic in certain doses or time periods. Some of the most common naturally occurring brain toxins that lead to neurotoxicity as a result of long term drug use are amyloid beta (Aβ), glutamate, dopamine, and oxygen radicals. When present in high concentrations, they can lead to neurotoxicity and death (apoptosis). Some of the symptoms that result from cell death include loss of motor control, cognitive deterioration and autonomic nervous system dysfunction. Additionally, neurotoxicity has been found to be a major cause of neurodegenerative diseases such as Alzheimer's disease (AD).

Neurotoxic agents

Amyloid beta

Amyloid beta (Aβ) was found to cause neurotoxicity and cell death in the brain when present in high concentrations. Aβ results from a mutation that occurs when protein chains are cut at the wrong locations, resulting in chains of different lengths that are unusable. Thus they are left in the brain until they are broken down, but if enough accumulate, they form plaques which are toxic to neurons. Aβ uses several routes in the central nervous system to cause cell death. An example is through the nicotinic acetylcholine receptor (nAchRs), which is a receptor commonly found along the surfaces of the cells that respond to nicotine stimulation, turning them on or off. Aβ was found manipulating the level of nicotine in the brain along with the MAP kinase, another signaling receptor, to cause cell death. Another chemical in the brain that Aβ regulates is JNK; this chemical halts the extracellular signal-regulated kinases (ERK) pathway, which normally functions as memory control in the brain. As a result, this memory favoring pathway is stopped, and the brain loses essential memory function. The loss of memory is a symptom of neurodegenerative disease, including AD. Another way Aβ causes cell death is through the phosphorylation of AKT; this occurs as the phosphate group is bound to several sites on the protein. This phosphorylation allows AKT to interact with BAD, a protein known to cause cell death. Thus an increase in Aβ results in an increase of the AKT/BAD complex, in turn stopping the action of the anti-apoptotic protein Bcl-2, which normally functions to stop cell death, causing accelerated neuron breakdown and the progression of AD.

Glutamate

Glutamate is a chemical found in the brain that poses a toxic threat to neurons when found in high concentrations. This concentration equilibrium is extremely delicate and is usually found in millimolar amounts extracellularly. When disturbed, an accumulation of glutamate occurs as a result of a mutation in the glutamate transporters, which act like pumps to clear glutamate from the synapse. This causes glutamate concentration to be several times higher in the blood than in the brain; in turn, the body must act to maintain equilibrium between the two concentrations by pumping the glutamate out of the bloodstream and into the neurons of the brain. In the event of a mutation, the glutamate transporters are unable to pump the glutamate back into the cells; thus a higher concentration accumulates at the glutamate receptors. This opens the ion channels, allowing calcium to enter the cell causing excitotoxicity. Glutamate results in cell death by turning on the N-methyl-D-aspartic acid receptors (NMDA); these receptors cause an increased release of calcium ions (Ca²⁺) into the cells. As a result, the increased concentration of Ca²⁺ directly increases the stress on mitochondria, resulting in excessive oxidative phosphorylation and production of reactive oxygen species (ROS) via the activation of nitric oxide synthase, ultimately leading to cell death. Aβ was also found aiding this route to neurotoxicity by enhancing neuron vulnerability to glutamate.

Oxygen radicals

The formation of oxygen radicals in the brain is achieved through the nitric oxide synthase (NOS) pathway. This reaction occurs as a response to an increase in the Ca²⁺ concentration inside a brain cell. This interaction between the Ca²⁺ and NOS results in the formation of the cofactor tetrahydrobiopterin (BH4), which then moves from the plasma membrane into the cytoplasm. As a final step, NOS is dephosphorylated yielding nitric oxide (NO), which accumulates in the brain, increasing its oxidative stress. There are several ROS, including superoxide, hydrogen peroxide and hydroxyl, all of which lead to neurotoxicity. Naturally, the body utilizes a defensive mechanism to diminish the fatal effects of the reactive species by employing certain enzymes to break down the ROS into small, benign molecules of simple oxygen and water. However, this breakdown of the ROS is not completely efficient; some reactive residues are left in the brain to accumulate, contributing to neurotoxicity and cell death. The brain is more vulnerable to oxidative stress than other organs, due to its low oxidative capacity. Because neurons are characterized as postmitotic cells, meaning that they live with accumulated damage over the years, accumulation of ROS is fatal. Thus, increased levels of ROS age neurons, which leads to accelerated neurodegenerative processes and ultimately the advancement of AD.

Dopaminergic Neurotoxicity

Endogenous

The endogenously produced autotoxin metabolite of dopamine, 3,4-Dihydroxyphenylacetaldehyde (DOPAL), is a potent inducer of programmed cell death (apoptosis) in dopaminergic neurons. DOPAL may play an important role in the pathology of Parkinson's disease.

Drug induced

Certain drugs, most famously the pesticide and metabolite MPP+ (1-methyl-4-phenylpyridin-1-ium) can induce Parkinson's disease by destroying dopaminergic neurons in the substantia nigra. MPP+ interacts with the electron transport chain in the mitochondria to generate reactive oxygen species which cause generalized oxidative damage and ultimately cell death. MPP+ is produced by monoamine oxidase B as a metabolite of MPTP (1-methyl-4-phenyl-1,2,3,6-tetrahydropyridine), and its toxicity is particularly significant to dopaminergic neurons because of an active transporter on those cells that bring it into the cytoplasm. The neurotoxicity of MPP+ was first investigated after MPTP was produced as a contaminant in the pethidine synthesized by a chemistry graduate student, who injected the contaminated drug and developed overt Parkinson's within weeks. Discovery of the mechanism of toxicity was an important advance in the study of Parkinson's disease, and the compound is now used to induce the disease in research animals.

Prognosis

The prognosis depends upon the length and degree of exposure and the severity of neurological injury. In some instances, exposure to neurotoxins or neurotoxicants can be fatal. In others, patients may survive but not fully recover. In other situations, many individuals recover completely after treatment.

The word neurotoxicity (/ˌnʊəroʊtɒkˈsɪsɪti/) uses combining forms of neuro- + tox- + -icity, yielding "nervous tissue poisoning".