Release Date: November 2006
Next Release Date: November 2007
New Reactor Designs
Reactor Design |
Vendor |
Approximate
Capacity (MWe) |
Reactor Type |
Certification Status |
Target Certification |
AP600 |
Westinghouse |
650 |
PWR |
Certified |
Certified |
AP1000* |
Westinghouse |
1117 |
PWR |
Certified |
Certified |
ABWR* |
GE et al |
1371 |
BWR |
Certified |
Certified |
System
80+ |
Westinghouse |
1300 |
PWR |
Certified |
Certified |
ESBWR* |
GE |
1550 |
BWR |
Undergoing certification |
2007 |
EPR* |
AREVA NP |
1600 |
PWR |
Pre-certification |
2009 |
PBMR |
Westinghouse, Eskom |
180 |
HTGR |
Pre-certification |
Not Available |
IRIS |
Westinghouse et al |
360 |
PWR |
Pre-certification |
2010 |
US
APWR |
Mitsubishi |
1600 |
PWR |
Undergoing certification |
2011 |
ACR
Series |
AECL |
700-1200 |
Modified PHWR |
Pre-certification |
Not Available |
GT-MHR |
General Atomics |
325 |
HTGR |
Research prototype planned |
Not Available |
4S* |
Toshiba |
10-50 |
Sodium-cooled |
Potential construction |
Not Available |
This document supersedes an earlier publication entitled “New
Reactor Designs”. Criteria for inclusion on the preceding
table are: Design certification issued by the Nuclear Regulatory Commission (AP600, AP1000,
APWR, System 80+)
1. Submission to the Nuclear Regulatory Commission (NRC) of an application
for design certification (ESBWR, USAPR)
2. Recent pre-design certification activities with the NRC or public announcement
of such intentions (EPR, PBMR, IRIS, ACR series reactors)
3. A research reactor design that has been discussed with the NRC that might
lead to a commercial prototype (GT-MHR)
4. Selected additional designs that appear to be intended for eventual construction
in the US. (4S)
Excluded are:
1. Reactors that do not appear to be intended for the US market.
2. Reactors that are components of US government programs that have not yet
been identified for targeted design certification. This excluded list
includes many designs associated with Generation IV (Gen IV) reactor designs
(included in the previous edition of “New Reactor Designs”),
the Next Generation Nuclear Power (NGNP) program, and the Global Nuclear
Energy Partnership (GNEP). Such reactor designs will be included after
the designs are publicly identified for design certification. Gen IV
reactors are summarized on http://nuclear.inl.gov/gen4/index.shtml.
Reactor Types
- Pressurized Water Reactors (PWR): PWRs use nuclear-fission to
heat water under pressure within the reactor. This water is then circulated
through a heat exchanger (called a "steam generator") where steam
is produced to drive an electric generator. The water used as a coolant
in the reactor and the water used to provide steam to the electric turbines
exists in separate closed loops that involve no substantial discharges
to the environment. Of the 104 fully licensed reactors in the United States,
69 are PWRs. Westinghouse, Babcock and Wilcox, and Combustion Engineering
designed the designed the nuclear steam supply systems (NSSS) for these
reactors. After these reactors were built, Westinghouse and Combustion
Engineering nuclear assets were combined. The French-German owned firm
Areva NP has acquired many of Babcock and Wilcox's nuclear technology rights,
though portions of the original Babcock and Wilcox firm still exist and
possess some nuclear technology rights as well. Other major makers of PWR
reactors, including Areva, Mitsubishi, and Russia’s
Atomstroyexport, have not yet sold their reactors in the U.S. www.eia.doe.gov/cneaf/nuclear/page/nuc_reactors/pwr.html
- Boiling Water Reactors (BWR): The remaining 35 operable reactors
in the United States are BWRs. BWRs allow fission-based heat from
the reactor core to boil the reactor’s coolant water into the steam
that is used to generate electricity. General Electric built all boiling
water reactors now operational in the United States. Areva NP and Westinghouse
BNFL have each designed BWRs.www.eia.doe.gov/cneaf/nuclear/page/nuc_reactors/bwr.html
- Pressurized Heavy Water Reactors (PHWR): PHWRs have been promoted
primarily in Canada and India, with additional commercial reactors operating
in South Korea, China, Romania, Pakistan, and Argentina. Canadian-designed
PHWRs are often called "CANDU" reactors. Siemens,
ABB (now part of Westinghouse), and Indian firms have also built commercial
PHWR reactors. Heavy water reactors now in commercial operation use
heavy water as moderators and coolants. The Canadian firm, Atomic
Energy of Canada Limited (AECL), has also recently proposed a modified
PHWR (the ACR series) which would only use heavy water as a moderator. Light
water would cool these reactors. No successful effort has been made
to license commercial PHWRs in the United States. PHWRs have been popular
in several countries because they use less expensive natural (not enriched)
uranium fuels and can be built and operated at competitive costs. The continuous
refueling process used in PHWRs has raised some proliferation concerns
because it is difficult for international inspectors to monitor. Additionally,
the relatively high Pu-239 content of PHWR spent fuel has also raised
proliferation concerns. The importance of these claims is challenged
by their manufacturers. PHWRs,
like most reactors, can use fuels other than uranium and the ACR series
of reactors is intended to use slightly enriched fuels. Particular
interest has been shown in India in thorium-based fuel cycles. http://www.eia.doe.gov/cneaf/nuclear/page/nuc_reactors/china/candu.html
- High Temperature Gas-cooled Reactors (HTGR): HTGRs are distinguished
from other gas-cooled reactors by the higher temperatures attained
within the reactor. Such higher temperatures might permit the reactor to
be used as an industrial heat source in addition to generating electricity. Among
the future uses for which HTGRs are being considered is the commercial
generation of hydrogen from water. In some cases, HTGR turbines run
directly by the gas that is used as a coolant. In other cases, steam
or alternative hot gases such as nitrogen are produced in a heat exchanger
to run the power generators. Recent proposals have favored helium
as the gas used as an HTGR coolant. The most famous U.S. HTGR example
was the Fort Saint Vrain reactor that operated between 1974 and 1989. Other
HTGRs have operated elsewhere, notably in Germany. Small research HTGR
prototypes presently exist in Japan and China. Commercial HTGR designs
are now promoted in China, South Africa, the United States, the Netherlands,
and France though none of these is yet commercially marketed. The
proposed Next Generation Nuclear Plant (NGNP) in the U.S. will most likely
be a helium-based HTGR, if it is funded to completion. http://www.nuc.berkeley.edu/designs/mhtgr/mhtgr.GIF
- Sodium-cooled reactors reactors: Sodium-cooled
reactors are included on this list primarily because of proposals
to build a Toshiba 4S reactor in Alaska. Sodium-cooled reactors
use the molten (liquid) metal sodium as a coolant to transfer reactor
generated heat to an electricity generation unit. Sodium-cooled
reactors are often associated with “fast
breeder reactors (FBRs)” though this is technically
not the case in the 4S design.
Links are provided solely as a service to our customers, and therefore should
not be construed as advocating or reflecting any position of the Energy Information
Administration (EIA). In addition, EIA does not guarantee the content
or accuracy of any information presented in linked sites.
AP600
(Westinghouse)
Synonyms: Advance Passive 600
Approximate Capacity (electric): 600 MWe
Reactor Type: Pressurized Water Reactor
NRC Design Certification Status: Certified December 1999
Supporting Generating Companies (potential site): None
The AP600 is a 600 MW PWR certified by the NRC. While based on previous PWR
designs, the AP600 has innovative passive safety features that permit a greatly
simplified reactor design. Simplification has reduced plant components and
should reduce construction costs. The AP600 has been bid overseas but has never
been built. Westinghouse has deemphasized the AP600 in favor of the larger,
though potentially even less expensive (on a cost per kilowatt or capacity
basis) AP1000 design.
Further Information: http://www.ap600.westinghousenuclear.com/ http://www.nei.org/index.asp?catnum=3&catid=704
AP1000
(Westinghouse)
Synonyms: Advanced Passive 1000
Approximate Capacity (electric): 1117-1154 MWe
Reactor Type: Pressurized Water Reactor
NRC Design Certification Status: Certified after December 2005, though amendments
have since been proposed.
Supporting Generating Companies (potential site): Duke Power (Cherokee County),
Progress Energy (Harris), Southern Company (Vogtle), NuStart Energy-Tennessee
Valley Authority (Bellefonte)
The AP1000 design is favored for construction at five to six potential sites
(ten to twelve reactors) in the United States. The AP1000 is an enlargement
of the AP600, designed to almost double the reactor's target electricity output
without proportionately increasing the total cost of building the reactor.
Westinghouse anticipates that operating costs should be below the average of
reactors now operating in the United States. While Westinghouse owns rights
to several other designs, the AP1000 is the principal product that the company
now promotes in the United States for near term deployment. The AP1000 includes
innovative, passive safety features and a much simplified design intended to
reduce the reactor’s material and construction costs while improving
operational safety. During 2007 or 2008 it is anticipated that the AP1000 will
be the subject of combined license (COL) applications to build and operate
new reactors in the United States. In early 2005 Westinghouse submitted
a bid to build a version of the AP1000 to build as many as four AP1000s at
two sites in China.
Further Information: http://www.nrc.gov/reactors/new-licensing/design-cert/ap1000.html http://www.ap1000.westinghousenuclear.com/ http://en.wikipedia.org/wiki/AP1000 http://www.nei.org/doc.asp?docid=770
ABWR
(General Electric and others)
Synonyms: Advanced Boiling Water Reactor
Approximate Capacity (electric): 1371-1465 MWe
Reactor Type: Boiling Water Reactor
NRC Design Certification Status: Certified May 1997. Design amendments
are possible but have not been publicly announced.
Supporting Generating Companies (potential site): NRG Energy (South Texas Project);
Amarillo Power
Four ABWRs operate in Japan and more are planned there. Two additional
ABWRs are under construction in Taiwan and two units are being considered for
the South Texas Project site in the United States. While the ABWR design is
usually associated in the United States with General Electric, variations on
the design have also been built by Toshiba and Hitachi. Hitachi also hopes
to associate with General Electric for building additional ABWRs at the South
Texas Project. The Tennessee Valley Authority (TVA) published a study
of the costs of building an ABWR reactor in the United States in September
2005 (below). Vendors now claim costs for building the ABWR that are low enough
that they have attracted some customer interest.
Further Information: http://www.gepower.com/prod_serv/products/nuclear_energy/en/new_reactors/abwr.htm http://en.wikipedia.org/wiki/ABWR http://www.nei.org/doc.asp?catnum=&catid=&docid=110&format=print http://np2010.ne.doe.gov/reports/Main%20Report%20All5.pdf http://www.nuc.berkeley.edu/designs/abwr/abwr.html
System 80+
(Westinghouse)
Synonyms: None
Approximate Capacity (electric): 1300 MWe plus
Reactor Type: Pressurized Water Reactor
NRC Design Certification Status: Certified May 1997.
Supporting Generating Companies (potential site): A modified version of the
design is being promoted for development in South Korea
The System 80+ reactor is a PWR designed by Combustion Engineering (CE) and
by CE's successor owners ABB and Westinghouse. The NRC has certified the System
80+ for the U.S. market, but Westinghouse no longer actively promotes the design
for domestic sale. The System 80+ provides a basis for the APR1400 design that
has been developed in Korea for future deployment and possible export.
Further Information: http://www.nei.org/index.asp?catnum=3&catid=703 http://www.nuc.berkeley.edu/designs/sys80/sys80.html
ESBWR
(General Electric)
Synonyms: Sometimes called Economic Simplified Boiling Water Reactor or
European Simplified Boiling Water Reactor though General Electric does not
frequently use the name.
Reactor Type: Boiling Water Reactor
Approximate Capacity (electric): 1550 MWe plus
NRC Design Certification Status: Undergoing certification
Supporting Generating Companies (potential site): Entergy (Grand Gulf, River
Bend), Dominion Energy (North Anna)
The ESBWR is a new simplified BWR design promoted by
General Electric and some allied firms. The ESBWR constitutes an evolution
and merging of several earlier designs including the ABWR. The ESBWR, which
includes new passive safety features, is intended to cut construction and operating
costs significantly from earlier ABWR designs. GE and others have invested
heavily in the ESBWR though the design and two US utilities, Dominion and Entergy
have expressed an interest in possibly building the design at three sites. These
utilities have stated that they might apply for a combined license (COL) to
build and operate new ESBWR reactors during 2007 or 2008. The two utilities
have also applied for Early Site Permits (ESPs) for the designs which the anticipate
receiving during 2007. The ESBWR is presently undergoing design certification
with the NRC.
Further Information: http://www.nrc.gov/reactors/new-licensing/design-cert/esbwr.html http://www.gepower.com/prod_serv/products/nuclear_energy/en/new_reactors/esbwr.htm http://en.wikipedia.org/wiki/ESBWR http://www.nei.org/index.asp?catnum=4&catid=907 www.ans.org/pubs/magazines/nn/docs/2006-1-3.pdf
EPR
(Areva NP)
Synonyms: Evolutionary Pressurized Water Reactor, the name European Pressurized
Water Reactor does not apply to the US design
Approximate Capacity (electric): 1600 MWe
Reactor Type: Pressurized Water Reactor
NRC Design Certification Status: Pre-application review
Supporting Generating Companies (potential site): UniStar Nuclear-Constellation-Areva
(Calvert Cliffs, Nine Mile Point)
Areva NP announced in early 2005 that it would market its EPR design in the
United States and has recently begun pre-certification activities. The
U.S.-market version is called the Evolutionary Pressurized Water Reactor. The
EPR is a conventional, though advanced, PWR in which components have been simplified
and considerable emphasis is placed on reactor safety. The design is now being
built in Finland with a target commercialization during 2010. The French government
has also authorized building an EPR at Flamanville 3 in France. Additional
EPRs might replace additional commercial reactors now operating in France starting
in the late 2010s and EPRs have been bid, in China and elsewhere. The
proposed size for the EPR has varied over time, but is most frequently placed
around 1600 MWe. Earlier designs were as large as 1750 MWe. The
EPR is promoted in the United States by UniStar Nuclear, a joint venture of
Constellation Energy and AREVA NP. UniStar is presently looking at the
possibility of building EPRs at Constellation-owned sites at Nine Mile Point
and Calvert Cliffs and has had discussions with other firms. Areva NP
anticipates submitting a design certification application to the Nuclear Regulatory
Commission during late 2007.
Further Information: http://www.nrc.gov/reactors/new-licensing/design-cert/epr.html http://en.wikipedia.org/wiki/European_Pressurized_Reactor http://unistarnuclear.com/
PBMR
(Westinghouse, PBMR Ltd.)
Synonyms: Pebble Bed Modular Reactor
Approximate Capacity (electric): 165 MWe
Reactor Type: High temperature gas-cooled reactor (HTGR)
NRC Design Certification Status: Pre-application review
Supporting Generating Companies (potential site): The design has no U.S. generating
company sponsor. The PBMR is supported by the South African utility Eskom
for development in South Africa
The PBMR uses helium as a coolant and is part of the HTGR family of reactors.
PBMR development is thus a product of a lengthy history of research, notably
in Germany and the United States. More recently the design has been promoted
and revised by PBMR Ltd., an affiliate of the South African utility Eskom.
Westinghouse is a minority investor in PBMR Ltd. and has taken a leading role
in U.S. design certification. The PBMR design is presently in a “pre-certification” status
with the NRC. Prototype variations on the PBMR design now operate in
China and Japan. Eskom has also received administrative approval to build
a prototype PBMR in South Africa. If the prototype is successful, Eskom
has stated it intends to build several follow on units. There is no U.S. generating
company sponsor of the design. At around 165 MWe the PBMR would be one
of the smaller reactors now proposed for the commercial market. This is considered
a marketing advantage by some because small reactors require lower initial
capital investments than larger new units. Several PBMRs could be built
at a single site as local power demand requires. The NRC also does not claim
the same familiarity with the PBMR design that it has with light water reactors
(PWR and BWR). Fuels used in the PBMR would be more highly enriched than
the uranium is now used in light water reactor designs. China and South Africa
have also discussed cooperation in PBMR efforts.
Further Information: http://www.nrc.gov/reactors/new-licensing/design-cert/pbmr.html http://www.pbmr.com/ http://en.wikipedia.org/wiki/Pebble_bed_modular_reactor http://www.nei.org/index.asp?catnum=3&catid=707
IRIS
(Westinghouse-led consortium)
Synonyms: International Reactor Innovative and Secure
Approximate Capacity (electric): 100-300 MWe
Reactor Type: Pressurized Water Reactor (advanced design)
NRC Design Certification Status (potential site): Pre-application review
Supporting Generating Companies: None, though international generating companies
are part of the international consortium developing the design.
Westinghouse has promoted the IRIS reactor design as a significant simplification
and innovation in PWR technology. While the IRIS is a PWR, several components,
notably steam generators, are internal to the reactor vessel. The reactor
design is smaller than most operating PWRs and is asserted to be much simplified.
Fuel for the IRIS would be more enriched (5-9% U-235 compared to 3-5%) than
is presently used in U.S. PWR. This might allow for longer periods between
reactor refueling. The IRIS reactor includes features intended to avoid
loss of coolant accidents. Pre-certification is proceeding though IRIS might
show its potential during the next decade (2010s). Certification activities
as now scheduled could precede commercial availability. IRIS sponsors
have a targeted 2010 certification completion date with commercial deployment
to follow.
Further Information: http://hulk.cesnef.polimi.it/ http://www.nei.org/index.asp?catnum=3&catid=712
US-APWR
(Mitsubishi Heavy Industries)
Synonyms: International Advanced Pressurized Water Reactor, the name Advanced
Pressurized Water Reactor (APWR) usually refers to the design in Japan
Approximate Capacity (electric): 1700 MWe in the United States
Reactor Type: Pressurized Water Reactor
NRC Design Certification Status: Pre-application review. Application
targeted for March 2008.
Supporting Generating Companies (potential site): Support exists for the related
APWR design among Japanese generating companies.
The US-APWR is a U.S.-marketed variation on APWR design sold in Japan by Mitsubishi
Heavy Industries. The 1538 MW APWR has been selected by Japan Atomic
Power Company for two units to be located at Tsuruga in Japan with the first
unit slated for completion in 2014. Other Japanese generating companies
are also interested in the APWR design. The 1700 MW US-APWR was only
recently (June 2006) announced for the U.S. market and is not presently being
certified in any other international markets. The US-APWR has not yet received
publicized support from any U.S. generating company. Pre-application
design certification activities before the U.S. Nuclear Regulatory Commission
began during July 2006. Mitsubishi targets a design certification application
for March 2008 and hopes complete the process during 2011. Mitsubishi
also wants to have the reactor available for construction in the U.S. as early
as 2011. Mitsubishi is also investigating certifying a second, smaller
reactor design at a capacity of 1200 MW.
Further Information: http://en.wikipedia.org/wiki/Advanced_Pressurized_Water_Reactor http://www.mhi-ir.jp/english/new/sec1/200607031122.html
ACR Series
(Atomic Energy of Canada Limited)
Synonyms: Advanced CANDU Reactor, ACR700, ACR1000
Approximate Capacity (electric): 700-1200 MWe
Reactor Type: Modified Pressurized Heavy Water Reactor
NRC Design Certification Status: Pre-application review apparently on hold.
Supporting Generating Companies(potential site): None, though it is among the
designs being considered for eventual development in Ontario, Canada.
AECL's ACR series of reactors is considered by its vendor to be an evolution
from the internationally successful CANDU line of PHWRs. Original pre-application
design certification procedures in the U.S. had been for the 700 MW ACR700
design. More recent discussions have focused on the 1200 MW ACR1000. CANDU
reactors and their Indian derivatives have had more success than any family
of commercial power reactors except the LWRs. One of the innovations in the
ACR series of reactors, compared to earlier CANDU designs, is that heavy water
is used only as a moderator in the reactor. Light water is used as the coolant.
Earlier CANDU designs used heavy water both as a moderator and as a coolant.
This change makes it debatable whether the ACR design series are true PHWRs,
PWRs, or a hybrid between the two designs. Fueling procedures for the
ACR follow the earlier CANDU designs in that it occurs while the reactors are
in service rather than during refueling outages. AECL has aggressively marketed
the ACR series offering low prices, short construction periods, and favorable
financial terms. As is the case for most non-LWR reactors, U.S. generating
companies, nuclear engineers, and regulators have only limited familiarity
with the design. Interest in the ACR series by Dominion Resources in Virginia
and by United Kingdom generating companies has not been sustained. AECL
has subsequently delayed its efforts to certify the design in the United States. The
ACR series has been mentioned as a possible contender for construction in Ontario,
the earliest possible reactor construction there might be either earlier CANDU
designs or non-Canadian designs.
Further Information: http://www.nrc.gov/reactors/new-licensing/design-cert/acr-700.html http://www.aecl.ca/Reactors/ACR-1000.htm http://www.aecl.ca/AssetFactory.aspx?did=88 http://en.wikipedia.org/wiki/Advanced_CANDU_Reactor
GT-MHR
(General Atomics)
Synonyms: Gas Turbine Modular Helium Reactor, Freedom Reactor (Entergy trademark)
Approximate Capacity (electric): 285 MWe
Reactor Type: High temperature gas-cooled reactor (HTGR)
NRC Design Certification Status: Pre-application review.
Supporting Generating Companies: Entergy (development only)
The GT-MHR is an HTGR developed by the U.S. firm, General Atomic. The most
advanced plans for GT-MHR development relate to building reactors in Russia
to assist in the disposal of surplus plutonium supplies. Parallel plans for
commercial power reactors would use uranium-based fuels enriched to as high
as 19.9 percent U-235 content. This would keep the fuel a fraction below the
20 percent U-235 enrichment that defines highly-enriched uranium. The U.S.
utility, Entergy, has participated in GT-MHR development and promotion and
uses the name "Freedom Reactor" for the design. A proposed
research version of the reactor has been proposed for the University of Texas
Permian Basin and affiliated institutions for Andrews County, Texas. Because
coolant temperatures arising from HTGRs are much higher than from light water
reactors, the design is viewed as a potential source of commercial heat. Particular
attention has been paid to the design's potential to produce of hydrogen from
water. The GT-MHR is considered, among many other designs, as a potential contender
for the US Department of Energy's Next Generation Nuclear Plant (NGNP) program.
Further Information: http://gt-mhr.ga.com/ http://en.wikipedia.org/wiki/GT-MHR http://www.nei.org/doc.asp?catnum=3&catid=711
4S
(Toshiba)
Synonyms: Super Safe, Small, and Simple
Approximate Capacity (electric): 10 MWe, larger possible
Reactor Type: Sodium-cooled
NRC Design Certification Status: Manufacturer and sponsor are developing a
pre-application approach.
Supporting Generating Companies (potential site): Town of Galena, Alaska
The 4S is a very small molten sodium-cooled reactor designed by Toshiba. The
reactor presently being considered is 10 MWe though larger and smaller versions
exist. The 4S is intended for use in remote locations and to operate
without refueling during its 30-year life. The 4S has been compared
with a nuclear “battery” because it does not require refueling. The
lack of refueling would mean that the reactor’s fuel supply would be a
capital cost rather than an operating cost. It has been suggested that
the fuel might be relatively low cost, reprocessed spent fuels originating
from more conventional power reactors. Other potential fuels are uranium
or uranium-plutonium alloys. If uranium is the fuel in the United States,
plans call for 19.9 percent fuel enrichment, just below the 20 percent definition
of highly enriched uranium. The use of molten-sodium as a coolant is not new,
having been used in many fast breeder reactors. Toward the end of 2004
the town of Galena, Alaska granted initial approval for Toshiba to investigate
building a 4S reactor in that remote location. The design is also under
consideration for other locations in Alaska. Most recent discussions
target completion around 2013, though the schedule is not firm. Galena
and Toshiba officials discussed their plans with the NRC in early February
2005 and plan additional filings over the coming years. The NRC indicated
that it was not familiar with the 4S design and that design certification (at
vendor expense) might be costly and prolonged. Design certification can
be incorporated in the COL process thus it is unclear if a separate design
certification will be pursued, if the project continues.
Further Information: http://en.wikipedia.org/wiki/Toshiba_4S http://www.atomicinsights.com/AI_03-20-05print.html http://www.iser.uaa.alaska.edu/Publications/Galena_power_draftfinal_15Dec2004.pdf#search='Toshiba
4S'
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