Overview
This article summarizes nuclear reactor designs that are
either available or anticipated to become available in the United States by
2030. Criteria for including reactors are: 1) participation or likely
participation in the U.S. Nuclear Regulatory Commission's design certification
or pre-certification programs; and 2) inclusion under the Generation IV
International Forum (GIF) program for longer-term reactor development. The U.S.
Department of Energy is among the sponsors of the GIF program. While no
detailed technical description of particular reactor designs is included, such
descriptions and schematics are available elsewhere and, when practical, some
of these are hyperlinked in the text. Reactor vendors who put forward new
designs anticipate that their designs will meet commercial market needs
including design safety and affordable, competitive construction costs while
maintaining the usually low operating costs of today’s commercial nuclear
reactors. This paper does not assess such views, though a section does identify
public discussion of efforts within the nuclear industry and the U.S.
government to improve the industry's competitive position.1
Existing
Reactor Designs and Design Categories
There are now 104 fully licensed nuclear power reactors in
the United States, though only 103 are now operational.2 Because each of these reactors is fully licensed and meets
national safety standards, a potential builder might choose to replicate any of
these designs for future construction. This is less likely, however, because
existing, operable reactors in the United States were initiated during or
before the 1970s. Technology has progressed and any future construction is
likely to incorporate more advanced designs intended to better meet today's
commercial and safety criteria.
There are possible exceptions to the preceding statement.
Four reactors in the United States were partially built and still possess valid
construction licenses. These reactors are WNP-1 in Washington State (Energy
Northwest), Watt's Bar 2 in Tennessee (Tennessee Valley Authority), and
Bellefonte 1 and 2 in Alabama (TVA). Moreover, these construction licenses have
been extended approximately to the end of the present decade. Construction on
each unit was halted almost two decades ago. Builders of these units, subject
to the rules of their licenses, have the right to resume construction on their
reactors that were designed during the 1970s or earlier. Whether the
construction under these existing designs will resume and whether former
designs will be continued remains to be determined, but appears unlikely. The
owners of WNP-1 have indicated an intention to forgo their construction license
to allow for eventual disassembly and clearance of present facilities.
All existing commercial nuclear reactors operating in the
United States fall into two broad categories, pressurized water reactor (PWR)
and boiling water reactor (BWR). Because both types of reactors are cooled and
moderated3 with ordinary "light" water, the two designs are
often grouped collectively as light water reactors (LWR). LWRs generate power
through steam turbines similar to those used for most power generated by
burning coal or fuel oil. Light water reactors have so far proven to be the
most commercially popular reactor design worldwide though there are notable
exceptions.4 There are several available websites that discuss existing
reactors in the United States. These include http://www.eia.doe.gov/cneaf/nuclear/page/nuc_reactors/reactsum.html.
Information on international operating reactors is available at http://www.iaea.org/programmes/a2.
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 exist 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 with
British Nuclear Fuels Limited to form Westinghouse BNFL. The French-German
owned firm Framatome ANP 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 Framatome ANP, Mitsubishi, and Russia’s
Atomstroyexport, have not yet sold their reactors in the U.S. A schematic
diagram of a PWR can be found at
http://www.eia.doe.gov/cneaf/nuclear/page/nuc_reactors/pwr.html.
The remaining 35 operable commercial nuclear 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. Framatome ANP and Westinghouse BNFL have each designed
BWRs. These have not yet been sold in the United States. A schematic diagram of
a BWR can be found at http://www.eia.doe.gov/cneaf/nuclear/page/nuc_reactors/bwr.html.
Although no LWR projects have been initiated in the United
States since the 1970s, the overall performance record of the existing fleet
has been reasonably successful. Some 111 LWRs have entered service in the U.S.
since 1969.5 Only seven of those have been permanently shut down. The
average annual capacity factor for nuclear reactors in the United States has
been around 90 percent during the 2000's. Average operating costs, as reported
by the Federal Energy Regulatory Commission, are slightly lower for LWRs than
for operating coal-fired plants and considerably below operating costs for
gas-fired plants.
Fuel costs for LWRs are particularly low.6
There have been attempts to operate additional classes of
reactors in the United States, though most examples were prototypes and were
not commercial successes. Perhaps the most famous example was the Fort Saint
Vrain reactor that operated between 1974 and 1989. It was a high temperature
gas-cooled reactor or HTGR. Other HTGRs operated elsewhere, notably in Germany.
HTGRs, of which there are many sub-categories, continue to stimulate commercial
interest. HTGR designs are promoted by firms in China, South Africa, the United
States, the Netherlands, and France.
There is some interest in building commercial HTGRs in several nations
including South Africa and China. Small
research prototypes already exist in Japan and China. HTGRs use a gas- helium
has been preferred- to generate electricity. In some cases the turbine is run
directly by the gas, in other cases steam or alternative hot gases such as
nitrogen are produced in a heat exchanger to generate the power. 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.
This improves HTGR’s suitability for commercial hydrogen production. Advocates
of HTGR designs hold that HTGRs have high safety, low costs, and a potential to
supply power to smaller markets than do LWRs.
HTGRs also are reputed to adapt better to changing load requirements of
electricity markets than LWRs.
Additional commercial reactor designs that operate outside
of the United States include fast breeder reactors (FBRs), pressurized heavy
water reactors (PHWRs), and gas-cooled reactors (GCRs). FBRs have received much
research funding but only limited market support. A "commercial" unit still operates in Russia and
prototypes exist elsewhere, notably France, Japan, and India. China also intends
to build a prototype FBR while India and Russia are building FBRs that might be
described as commercial.
"Breeder" or "fast" reactors have advantages because
U-235 is the only naturally occurring uranium isotope that is directly suitable
for commercial energy production. U-235 is only 0.7 percent of natural uranium.7 Most natural uranium is the U-238 isotope that is not
directly usable as a reactor fuel. During the course of any reactor’s operation
a portion of the U-238 in the fuel is converted to plutonium, primarily the
useful Pu-239 isotope, which provides a large portion of the energy used in
nuclear power production. The bulk of the U-238 content in a commercial reactor
is typically not converted to plutonium nor does it contribute significantly to
electricity production. A breeder
reactor converts more U-238 to usable fuels than the reactor consumes. Any
unused fuel produced by this procedure would have to be "reprocessed"
before some of the plutonium and the remaining U-235 and U-238 might again be
usable as a reactor fuel. FBRs have, so far, proven to be more expensive to
build and operate than LWRs. It is unclear whether this is because most FBRs
have been prototypes or if this reflects underlying costs. The plutonium
content of the spent and reprocessed fuel also raises concerns over weapons
proliferation. Many earliest FBR designs experienced system failures, though
some, notably the BN-600 in Russia, have operated reliably over extended
periods. Proponents of advanced reactor designs believe that some commercial
FBR designs could be deployed prior to other highly advanced, though untested
reactor designs.8
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.9 Siemens, ABB (now part of Westinghouse), and Indian firms
have also built commercial PHWR reactors. Commercial heavy water reactors now
in operation use heavy water as moderators and coolants. No successful effort
has been made to license PHWRs in the United States. PHWRs have proven to be
popular in several countries because they use less expensive natural (not
enriched) uranium fuels and can be built and operated at competitive costs.
PHWRs have often been preferred by nations wishing to develop an indigenous
fuel cycle without expensive enrichment facilities. The continuous process of
refueling PHWRs have raised some proliferation concerns as has the high Pu-239
content of the spent fuel. PHWRs, like most reactors, can use fuels other than uranium. Particular interest has been shown in
thorium-based fuel cycles.10
The term gas-cooled reactor (GCR) can be used ambiguously.
HTGRs, for example, are a subset of GCRs that operate at higher temperatures.
As used here, GCRs include "Magnox" reactors designed and built in
the United Kingdom since the 1950s and the derivative, advanced gas-cooled
reactor (AGR), also operated in the United Kingdom. Similar reactors had been
built and operated in France, Sweden, and Japan but have since closed. No GCR
design, as defined here, has operated commercially in the United States.
Commercial GCRs11 in the United Kingdom have operated longer than any
category of commercial reactors anywhere else in the world. Like the PHWRs, the
original GCR designs use natural uranium fuels, though newer designs (AGRs) use
slightly enriched fuels and are not confined to uranium fuels.12
Other potential designs for commercial reactors abound.
They have not been widely or recently considered for commercial applications in
the United States. There is some experience with additional concepts elsewhere
and at research facilities.
New Designs
1. Certified Designs
In recent years, the Nuclear Regulatory Commission (NRC)
has set up a process by which reactor designs might be certified prior to any
actual construction plans. The certification process seeks to reduce site
development time by resolving design issues prior to construction. Design
certification is an optional process and might occur simultaneously with site
licensing or construction licensing.
Normally reactor certification is the responsibility of the reactor
vendor rather than any utility that might choose to build a new reactor.
Certification Process for New
Reactors in the United States
|
Reactor Design
|
Lead Vendor(s)
|
Design Category
|
Status at NRC
|
System 80+
|
Westinghouse BNFL
|
PWR
|
Certified
|
ABWR
|
GE, Toshiba, Hitachi
|
BWR
|
Certified
|
AP600
|
Westinghouse BNFL
|
PWR
|
Certified
|
AP1000
|
Westinghouse BNFL
|
PWR
|
Finalizing Certification
|
ESBWR
|
GE
|
BWR
|
Pre-certification
|
SWR-1000
|
Framatome ANP
|
BWR
|
Pre-certification, deferred
|
ACR700
|
AECL
|
PHWR/PWR hybrid
|
Pre-certification
|
PBMR
|
Eskom
|
HTGR
|
Pre-certification, deferred
|
GT-MHR
|
General Atomic
|
HTGR
|
Pre-certification
|
IRIS
|
Westinghouse BNFL
|
PWR
|
Pre-certification
|
EPR
|
Framatome ANP
|
PWR
|
Pre-certification
|
ACR1000
|
AECL
|
PHWR/PWR hybrid
|
No application decision
|
4S
|
Toshiba
|
Sodium-cooled
|
No application decision
|
Any new reactor built in the United States over the next
decade or so would most likely use designs either recently certified by the NRC
or that will be certified by the NRC in the near future. (Design approval can
alternatively coincide with construction and operation licensing, skipping the
certification process.) The re-creation
of older designs is popular overseas and cannot be ruled out in the United
States. Presently there are three certified new reactor designs in the United
States: the System 80+, the Advanced Boiling Water Reactors (ABWR), and the
AP600. These designs are sometimes called Advanced Light Water Reactors (ALWR)
because they incorporate more advanced safety concepts than the reactors
previously offered by vendors. They are also sometimes called Generation III
reactors to distinguish them from earlier designs now operating in the U.S. and
globally and from later designs now seeking certification which are sometimes
called Generation III plus. Design certifications can expire if not supported
by a vendor.
System 80+ (Westinghouse BNFL): The System 80+ reactor is a PWR
designed by Combustion Engineering (CE) and by CE's successor owners ABB and
Westinghouse BNFL. The NRC has certified the System 80+ for the U.S. market,
but Westinghouse BNFL no longer actively promotes the design for domestic sale.
The System 80+ provides the basis for the APR1400 design that has been
developed in Korea for future deployment and possible export. Information on
the System 80+ reactor can be found on http://www.nei.org/index.asp?catnum=3&catid=703
and http://www.nuc.berkeley.edu/designs/sys80/sys80.html.
ABWR (General Electric, Toshiba, Hitachi): Among the three NRC-
certified ALWR designs only the ABWR has been deployed. Three ABWRs operate in
Japan, and three are under construction, two in Taiwan and one in Japan. While
the ABWR design is usually associated in the United States with General
Electric, units now being built in Japan are products of Toshiba and Hitachi.
Toshiba, and Hitachi frequently associate with General Electric in possible ABWR
projects in the U.S. There are many
variations in ABWR design. The most frequently mentioned capacities are in the
1250-1500 MWe range though smaller and larger designs have been proposed
depending on the vendor. Vendors now claim costs for building the ABWR that are
low enough that they have attracted some customer interest. Information on the ABWR can be found at http://www.nei.org/doc.asp?docid=110,
and http://www.nuc.berkeley.edu/designs/abwr/abwr.html
AP600 (Westinghouse BNFL): The AP600 is a 600 MW PWR certified
by the NRC. The AP600, while based on previous PWR designs, 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 less
expensive (on a kilowatt basis) AP1000 design. Information on the AP600 can be
found at http://www.ap600.westinghousenuclear.com/
and at http://www.nei.org/index.asp?catnum=3&catid=704.
The initial ALWR designs as a group have been praised for
their improvements in reactor safety and simplicity, but construction costs on
a “per kilowatt of capacity” basis might remain a barrier to commercial success
in the U.S. The ABWR design however has many variations and continues to be
selectively promoted by several vendors. It has been evaluated, along with
other designs, for construction at Bellefonte by the Tennessee Valley Authority
(TVA).
2. Undergoing Certification
Only one reactor design, the AP1000, is presently
undergoing certification with the NRC. This situation could change shortly as
additional designs move from "pre-certification" to actual
"certification". The certification process is anticipated to begin
for several additional designs during 2005 and 2006. Designs that vendors anticipate submitting for certification
during the next two years include the ESBWR, the ACR700, the EPR and IRIS. The process of certification takes several
years and depends heavily on how unique the proposed design is and whether the
design is supported by potential vendors and buyers. NRC hearings have emphasized that new and innovative designs
might take more time for certification because of limited NRC staff familiarity
with the designs.
AP100013 (Westinghouse BNFL): Quite often when a reactor is named,
its name includes digits such as the "1000" in the AP1000. This
usually indicates the initial electricity generating capacity of the design, in
this case 1000 MWe. Seldom do the digits indicate the present design capacity
as the design evolves. The most recent AP1000 design has been bid in China with
a 1175 MW-capacity. The AP1000 is an enlargement of the AP600, designed to
almost double the reactor's target output without proportionately increasing
the total cost of building the reactor. Westinghouse anticipates that operating
costs are anticipated to be below the average of reactors now operating in the
United States. While Westinghouse BNFL owns rights to several other designs,
the AP1000 is the principal product that the company now promotes in the United
States for near term construction. The AP1000 is a PWR with innovative, passive
safety features and a much simplified design intended to reduce the reactor’s
material and construction costs while improving operational safety. One
consortium of nine utilities called NuStart Energy promotes the AP1000 in the
United States and has informed the NRC that it intends to apply for a combined
construction and operating license (COL) for the design. This is not a commitment to build the
design. Westinghouse submitted a bid in early 2005 to build as many as four
AP1000s at two sites in China.
Information on the AP1000 can be found at http://www.nei.org/doc.asp?docid=770.
Information related to NRC certification for the AP1000 can be found athttp://www.nrc.gov/reactors/new-licensing/design-cert/ap1000.html.
3. Undergoing Pre-Certification
While pre-certification is a technical concept within the
NRC regulatory environment, the process can mean many things to potential
reactor vendors. Concepts such as the ESBWR, and the ACR700 appear to be much
further along toward certification than the other designs.14 The French designed EPR is undergoing construction in
Finland and has recently moved to pre-certification. Pre-certification represents a vendor's intention to proceed
toward commercialization in the U.S. and perhaps globally. Pre-certification is
a less expensive early stage of the certification process. Actual certification
procedures are much more complex. At an
early stage in pre-certification most NRC regulatory costs are borne by the
applicant.
ESBWR (Economic Simplified, Boiling Water Reactor) (General
Electric): The ESBWR15 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 that are now less actively
pursued by GE and other vendors beyond the exceptional case of Bellefonte in
Alabama. The intent of the new design, which includes new passive safety
features, is to cut construction and operating costs significantly from earlier
ABWR designs. GE and others are investing heavily in the ESBWR though the
design might not be available for deployment for several years. The ESBWR’s builders however anticipate that
the design will be available in time to meet any potential construction targets
in the U.S. The nine-utility NuStart Energy group promotes the ESBWR as well as
the AP1000 design. NuStart has informed
the NRC that it intends to apply for a COL for the ESBWR in addition to any
AP1000 application. Dominion Resources
is also evaluating the ESBWR for its North Anna plant in Virginia but has not
declared its COL intentions for the design.
Information related to certification of the ESBWR can be found at http://www.nrc.gov/reactors/new-licensing/design-cert/esbwr.html.
Siedewasser Reaktor (SWR-1000) (Framatome ANP): The SWR-1000 is a Framatome
ANP design for an advanced BWR. Framatome ANP was created through the merger of
the French nuclear vendor Framatome and the nuclear power assets of the German
firm Siemens. The SWR-1000 was originally designed by Siemens. Framatome ANP
began SWR-1000 pre-certification with the NRC several years ago. The SWR-1000
presently has no U.S. utility sponsor and is no longer being actively promoted
by Framatome which now emphasizes its EPR design. Literature on the design
notes the reactor's passive safety features. Passive safety also potentially
mean lower construction costs though this has not been as heavily promoted by
Framatome. Information on the SWR1000 can be found on http://www.de.framatome-anp.com/anp/e/foa/anp/products/s112.htm.
Information related to certification of the SWR-1000 can be found at http://www.nrc.gov/reactors/new-licensing/design-cert/swr-1000.html.
ACR700 (Atomic Energy of Canada Limited): AECL's "Advanced
CANDU Reactor" ACR70016 has been developed over a lengthy period of time and is
considered by its vendor to be an evolution from AECL's internationally
successful CANDU line of PHWRs. CANDU reactors and their Indian derivatives
have been more of a commercial success than any other line of power reactors
except the LWRs. One of the innovations in the ACR700, 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
ACR700 is a PHWR, a PWR, or a hybrid between the two designs. AECL has
aggressively marketed the ACR700 offering low prices, short construction
periods, and favorable financial terms. As is the case for most non-LWR
reactors, most U.S. utilities, nuclear engineers, and regulators have only
limited working familiarity with the design. Interest was initially shown by
Dominion Resources regarding possible construction at North Anna (Virginia) as
well as by utilities in several international locations, notably in Canada and
the United Kingdom. Dominion has recently switched to the ESBWR design for
North Anna in anticipation of the slow regulatory approval process for the
innovative Canadian-design. AECL has
subsequently slowed its efforts to certify the ACR700 in the United States
though the firm still intends to begin the certification process toward the end
of 2005. AECL announcements indicate
increased interest in a larger ACR1000 design.
Information on the ACR700 can be found on http://www.aecltechnologies.com/Content/ACR/default.htm
and http://www.aecl.ca/index.asp.
Information related to certification of the ACR-700 can be found at http://www.nrc.gov/reactors/new-licensing/design-cert/acr-700.html.
Pebble-bed Modular Reactor (PBMR) (Eskom): The PBMR, which uses helium as a
coolant, is part of the HTGR family of reactors and 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 the South African utility Eskom and
its affiliates. Westinghouse BNFL is a minority investor. Prototype variations of the PBMR are now
operating in China and Japan. Eskom has
received administrative approval to build a prototype PBMR in South Africa, but
has also been delayed in implementation by judicial rulings regarding the
reactor’s potential environmental impact. Certification procedures in the U.S.
have slowed, but never have been abandoned. At around 165 MWe the PBMR is one
of the smallest reactors now proposed for the commercial market. This is
considered a marketing advantage because new small reactors require lower
capital investments than larger new units.
Several PBMRs might be built at a single site as local power demand
requires. Small size has been viewed as a regulatory disadvantage because most
licensing regulations (at least formerly) required separate licenses for each
unit at a site. The NRC also does not claim the same familiarity with the
design that it has with LWRs. Fuels
used in the PBMR would include more highly enriched uranium than is now used in
LWR designs. The PBMR design is considered a possible contender for the U.S.
Department of Energy's Next Generation Nuclear Plant (NGNP) program in Idaho.
China has also indicated interest in building its own variation of the PBMR.
China and South Africa have also discussed cooperation in their efforts. Details regarding the PBMR design can be
found on https://www.pbmr.com/. Information
related to certification of the PBMR can be found at http://www.nrc.gov/reactors/new-licensing/design-cert/pbmr.html.
Gas-turbine Modular Helium
Reactor (GT-MHR) (General Atomic):
The GT-MHR is an HTGR design developed primarily 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 just below the
20 percent enrichment that defines highly enriched uranium. In initial GT-MHR
designs, the conversion of the energy to electricity would involve sending the
heated helium coolant directly to a gas turbine. There has been concern
regarding untested, though non-nuclear aspects of this generation process. This
has led potential sponsors to advocate similar ideas involving less innovative
heat transfer mechanisms prior to generating electricity or commercial heat.
The U.S. utility, Entergy, has participated in GT-MHR development and promotion
and has used the name "Freedom Reactor" for the design. Because
coolant temperatures arising from HTGRs are much higher than from LWRs, the
design is viewed as an improved commercial heat source. There has been
particular attention paid to the design's potential in the production of
hydrogen from water. The GT-MHR is considered a potential contender for the US
Department of Energy's Next Generation Nuclear Plant (NGNP) program.
Information on the GT-MHR can be found on
http://www.ga.com/gtmhr/. Information related to certification of the
GT-MHR can be found athttp://www.nrc.gov/reactors/new-licensing/design-cert/gt-mhr.html.
International Reactor
Innovative and Secure (IRIS) (Westinghouse
BNFL led consortium): Westinghouse BNFL has promoted the IRIS reactor design as a significant
simplification and innovation in PWR technology. The reactor design is smaller
than most operating PWRs and would be much simplified. The IRIS reactor
includes features intended to avoid loss of coolant accidents.
Pre-certification is proceeding. The IRIS
reactor may show
potential during the next decade.
Certification could precede commercial availability. IRIS has a targeted 2010 certification completion
date. IRIS presently has no utility sponsor in the U.S. Information on the IRIS can be
found on http://www.nei.org/index.asp?catnum=3&catid=712.
Information related to certification of the IRIS can be found at http://www.nrc.gov/reactors/new-licensing/design-cert/iris.html.
European Pressurized Water Reactor (EPR) (Framatome ANP): Framatome ANP announced in early
2005 that it would market its EPR design in the United States and has recently
begun pre-certification. 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 completion during 2009. The French government
also proposes building an additional EPR at Flamanville 3 in France. Present French policy suggests that additional
EPRs might replace additional commercial reactors now operating in France
starting in the late 2010s. The EPR was bid in early 2005 in competition to the
AP1000 for four reactors at two sites in China. The proposed size for the EPR has varied considerably over time
but might be around 1600 MWe. Earlier
designs were as large as 1750 MWe. In
either case the EPR would be the largest design now under consideration in the
United States. Some redesign might occur for the U.S. market. Framatome had earlier
indicated that U.S. certification for the EPR would occur after European
development proceeded. This decision
has since been made and the U.S. utility Duke Power is evaluating the EPR,
along with the AP1000 and ESBWR, for a COL application process that began
during 2005. A formal COL application
by Duke would occur several years later though design selection might occur
earlier. Framatome has posted material
on the EPR on
http://www.framatome-anp.com/servlet/ContentServer?pagename=Framatome-ANP%2Fview&c=rubrique&cid=1049449651371&id=1049449651371.The
NRC has not yet posted a status page for the EPR but one might be anticipated
on http://www.nrc.gov/reactors/new-licensing/design-cert.html.
4. Anticipated for Possible
Pre-Certification
Two designs, the ACR1000 and the 4S have not been formally
submitted for pre-certification in the United States. Because of the attention
that the designs are now receiving and their potential submission for
certification, they are summarized below.
ACR1000 (Atomic Energy of Canada Limited): While AECL has promoted
its ACR700 design, an ACR1000 has been designed as well. If the scale economies attributed by
Westinghouse BNFL to its AP series and by GE's its ABWR/ESBWR series are valid,
one might anticipate parallel, cost-lowering results for the ACR series.
Advertised costs for the ACR700 are already as low as any design proposed in
the United States for the near term. Promised construction times, as short as
three years, would set modern records for large reactor completion. When
Dominion Resources indicated in late 2004 that it was no longer pursuing ACR700
construction at North Anna, AECL stated that while it will continue with ACR700
certification, perhaps in late 2005, more effort would be placed on the 1100+
MWe ACR1000 design. Information on the ACR1000 can be found on http://www.aecl.ca/index.asp?menuid=21&miid=519&layid=3&csid=294.
4S (Toshiba): 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 designed for use in
remote locations and to operate for decades without refueling. This has led to the reactor to be compared
with a nuclear “battery”. The use of molten-sodium as a coolant is not
particularly new, having been used in many FBR designs. Sodium-coolants allow for higher reactor
temperatures. Potential fuels are uranium
or uranium-plutonium alloys. When
uranium is the likely fuel in the United States, present plans call for 19.9
percent fuel enrichment. This high
level of enrichment is one reason the reactor could be able to operate for
extended periods without refueling.
Toward the end of 2004 the town of Galena, Alaska granted initial approval
for Toshiba to build a 4S reactor in that remote location. Original plans called for completion in 2010
though it was acknowledged that this was ambitious. Galena and Toshiba officials discussed their plans with the NRC
in early February 2005. 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 not clear if a separate design certification will be
pursued, if the project continues. A
University of Alaska study of the proposed Galena reactor is available on http://www.iser.uaa.alaska.edu/Publications/Galena_power_draftfinal_15Dec2004.pdf#search='Toshiba%204S'
5. Generation IV (Gen IV) Concepts
The U.S. Department of Energy participates in the
Generation IV International Forum (GIF), an association of thirteen nations
that seek to develop a new generation of commercial nuclear reactor designs
before 2030. The U.S., Canada, France,
Japan and the United Kingdom signed an agreement on February 28, 2005 for
additional collaborative research and development of Gen IV systems. Criteria
for inclusion of a reactor design for consideration by the initial GIF group
include:
1.
Sustainable energy (extended fuel availability, positive environmental impact);
2. Competitive energy (low costs,
short construction times);
3. Safe and reliable systems
(inherent safety features, public confidence in nuclear energy safety); and
4. Proliferation resistance (does not
add unduly to unsecured nuclear material) and physical protection (secure
from terrorist attacks).
GIF members agreed during 2002 to concentrate their efforts
and funds on six concept designs whose goal is to become commercially viable
between 2015 and 2025. There is thus some leeway between the 2030 target for
the GIF program implementation and the targets for individual concepts.
Individual GIF participant nations are free to pursue any individual technology
they choose. The United States intends to pursue each design.
The GIF group, along with the U.S. Department of Energy's
Nuclear Energy Research Advisory Committee (NERAC), published "A
Technological Roadmap for Generation IV Nuclear Energy Systems" (December
2002) which summarizes plans and designs for Generation IV projects. This is
accessible through http://gif.inel.gov/roadmap/pdfs/gen_iv_roadmap.pdf
and describes each design in some detail including reactor schematics. Each
design is evolutionary; thus while the following descriptions involve
comparison to present designs, these analogies should be interpreted with
caution. Designs are expected to evolve. Gen IV programs are summarized on http://www.inel.gov/initiatives/generation.shtml.
The U.S. Department of Energy and the Idaho National
Laboratory are developing a program, the Next Generation Nuclear Plant (NGNP),
for implementing the first Gen IV reactor designs, and have initiated
discussions with potential private managers of the project. Potential portions of this program are
included in the above discussion of the GT-MHR and PBMR designs above. The NGNP program anticipates completing the
first Gen IV concept by 2020 and possibly earlier. Project efforts will include the production of hydrogen at the
prototype reactor. While very high
temperature gas-cooled reactors appear most likely for eventual consideration,
additional U.S.-based Gen IV designs might be submitted to the program
managers.
Nuclear Regulatory Commission officials have indicated that
present staff at the NRC are not familiar with innovative reactor designs, thus
any application for design certification would consume more time than for more
evolutionary LWR designs. Because GIF
reactors involve very long term plans, NRC familiarity with designs might evolve
before Generation IV reactors are ready for design certification.
Gas-cooled Fast Reactor (GFR): The GFR uses helium coolant directed to a gas
turbine generator to produce electricity. This parallels PBMR and original
GT-MHR designs. The primary difference from these designs is that the GFR would
be a "fast" or breeder reactor. One favored aspect of the design is
that it would minimize the production of many undesirable spent fuel waste
streams. The reference design size was targeted to be 288 MWe with a deployment
target date of 2025. In addition to producing electricity the design might be
used as a process heat source in the production of hydrogen. For further information see http://nuclear.inl.gov/gen4/gfr.shtml
Lead-cooled Fast Reactor (LFR): So far, most breeder reactors have used molten
metal technologies for their coolants. Many FBRs have used molten sodium, a
metal with which there is considerable experience but which has sometimes proven
difficult to handle. The LFR uses molten lead or a lead-bismuth alloy as its
coolant. Similar designs are being investigated in Russia which is not a GIF
participant. Some designs favored under the Generation IV program would result
in long periods between refueling, as much as 20 years or more. Target ranges
for this reactor would be 50-150 MWe. That would be rather small by historic
nuclear standards, but might meet localized market needs. Designs as large as
1200 MWe have been suggested. Initial targeted deployment would be in 2025.
Proposed designs would favor electricity production though proponents consider
the production of process heat at LFRs as possible. For further information see
http://nuclear.inl.gov/gen4/lfr.shtml. One design in this family of reactors is
described on http://www.coe.berkeley.edu/labnotes/1002/reactor.html.
Molten Salt Reactor (MSR): The MSR involves a circulating liquid of sodium,
zirconium, and uranium fluorides as a reactor fuel though the design could use
a wide variety of fuel cycles. The MSR has been presented as providing a
comparatively thorough fuel burn, safe operation, and proliferation resistance.
The initial reference design would be 1000 MWe with a deployment target date of
2025. Temperatures would not be as hot as for some other advanced reactors, but
some process heat potential exists. Versions of the MSR have been around for
some time but were never commercially implemented. The MSR was down rated
within the Gen IV program during 2003 because it was seen as too distant in the
future for inclusion within the Gen IV schedule. At the same time proponents see some MSR potential for the NGNP program. For further information see http://nuclear.inl.gov/gen4/msr.shtml.
Sodium-cooled Fast Reactor (SFR): Sodium-cooled fast reactors have been the most
popular design for breeder reactors. Designs have been proposed under the
Department of Energy’s “roadmap" for Generation IV reactors ranging from
150 to 1700 MWe. Elements of the SFR are included in the 4S design proposed by
Toshiba for Galena, Alaska. Molten
metal technology is no longer "new" but several early SFR prototypes
had difficulty obtaining sustained operation.
The BN-600 in Russia has been regarded as highly reliable. Design
supporters believe that the SFR promises superior fuel management characteristics.
The original target deployment date of 2015 reflected the considerable research
that the design has already received though the design is clearly not as ready
for U.S. deployment as LWR designs being evaluated for roughly the same period.
The target date seems to be lagging as the VHTR designs gain favor. Prototypes
have been built in France, Japan, Germany, the United Kingdom, Russia, India,
and the United States starting as early as 1951. Initial deployment would
probably focus on electricity due to comparatively low "outlet temperatures"
for the design. Sodium-cooled reactors are discussed at http://nuclear.inl.gov/gen4/sfr.shtml
and http://www.nuc.berkeley.edu/~gav/almr/01.intro.html.
Supercritical-water-cooled
Reactor (SCWR): The SCWR design is to be
the next step in LWR development and has been proposed with alternatives that
evolve from both the BWR and the PWR. SCWRs would operate at higher
temperatures and thermal efficiencies than present LWRs. The reference plant
might be 1700 MWe, at the upper end of present LWR designs. The deployment
target date was 2025. Some GIF participants favor the SCWR design because it is
more familiar to commercial markets than are more innovative concepts. Much of
the design research has been in Japan. Designers intend the SCWR to be much
less expensive to build than today's LWRs though some of these economies appear
to be shared by units now undergoing certification or pre-certification.
Operating cost savings are also anticipated. For further information see http://nuclear.inl.gov/gen4/scwr.shtml.
Very-high-temperature Reactor (VHTR): The VHTR is an evolution from the HTGR family of
reactors but would operate at even higher temperatures than designs now
undergoing pre-certification. Some of
the VHTR design standards might be met by modified PBMRs or GT-MHRs. In contrast with the GFR, the VHTR would not
be a breeder reactor, thus it would produce less potentially usable fuel than
it consumes. In addition to generating electricity, the design can provide
process heat for industrial activities including hydrogen production and
desalinization. Deployment is targeted for 2020, earlier than most Generation
IV designs. The VHTR is now a favored design in the U.S., where it is the basis
for most anticipated submissions for the still-evolving Next Generation Nuclear
Plant (NGNP). France also favors the design which is also popular in Asia and
South Africa. The VHTR is discussed at http://nuclear.inl.gov/gen4/vhtr.shtml
and http://www.nuc.berkeley.edu/designs/mhtgr/mhtgr.html.
Each GIF project involves new or untested reactor design
concepts. It would be surprising if each design concept met the program's
initial targets or that prototypes would match originally intended standards.
The research involved in the program has the potential to contribute to the
understanding of alternative types of commercial nuclear power and process heat
production even if individual projects fail to meet initial expectations.
6. Outlook
Efficiency Issues
A primary source of doubt regarding the potential of
nuclear power, at least in the U.S., has been whether the recent nuclear
technology has been too expensive to compete in the commercial marketplace.
There have been no orders for new nuclear power plants during the last three
decades in the United States and Canada. Finland’s order for a new reactor in
2003 broke a similar extended hiatus in Western Europe, excepting France where
orders tailed off later. France now looks likely to follow. Reactor vendors
have not ignored the message that their product has recently involved high
investment costs and long construction periods. Vendors now seek to position
their product with promises of lower prices, shorter construction times, and
specified financial arrangements. Most competitors are now offering fixed and
historically low prices for at least the nuclear components of their
designs. These promises vary with the
price of basic materials such as steel and concrete and as first of a kind engineering
costs are allocated or eliminated.
Location, buyer specifications, and regulatory requirements can also
alter anticipated costs.
Concerns regarding construction costs for new nuclear power
plants contrast sharply with the comparatively low cost of operating commercial
reactor designs. Overall operating costs for nuclear power plants, as reported
to the Federal Energy Regulatory Commission (FERC), have been roughly the same
as (most recently slightly less than) operating costs for coal-fired plants for
about two decades. Such operating costs are considerably below the costs of
operating most natural gas-fired generation units even when natural gas prices
are relatively low. Moreover, the fuel cost component of operating a nuclear
power plant is particularly low. This operating cost advantage has given
existing nuclear power units a favored position in the provision of base load
electric power. Nuclear plant designers hope to take advantage of such low
operating costs in positioning their new designs. Whether they will succeed has not yet been demonstrated. Discussions of estimates of the capital and
operating cost of new power generation units can be found in the "Issues
in Focus" section of the Annual Energy Outlook 2004 and in the Electricity Module of
the Assumptions for the Assumptions
for the Annual Energy Outlook 2005.
The following publications summarize efforts and procedures
to make new nuclear power plants commercially attractive.
- "Strategies for
competitive nuclear power plants (TECDOC-1123)" International Atomic
Energy Agency (November 1999).
- "A Roadmap to
Deploy New Nuclear Power Plants in the United States by 2010,": http://www.nuclear.gov/NucPwr2010/NucPwr2010_PI.html.
- Scully Capital,
"Final Draft, Business Case for Nuclear Power Plants, Bringing Public
and Private Resources Together for Nuclear Energy" (July 2002): http://www.nuclear.gov/home/bc/businesscase.html
- "A Technology
Roadmap for Generation IV Nuclear Energy Systems (GIF-002-00)": http://gif.inel.gov/roadmap/pdfs/gen_iv_roadmap.pdf
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