To explore these opportunities, the U.S. Department of Energy's
Office of Nuclear Energy, Science and Technology has engaged
governments, industry, and the research community worldwide in a
wide-ranging discussion on the development of next-generation
nuclear energy systems known as "Generation IV".
This has resulted in the formation of the Generation-IV
International Forum (GIF), a group whose member countries are
interested in jointly defining the future of nuclear energy
research and development.
In short, "Generation IV" refers to the development
and demonstration of one or more Generation IV nuclear energy
systems that offer advantages in the areas of economics, safety
and reliability, sustainability, and could be deployed
commercially by 2030.
A Generation IV Technology Roadmap is being prepared by GIF
member countries which will identify the six to eight most
promising reactor system and fuel cycle concepts and the R&D
necessary to advance these concepts for potential
commercialization. The Roadmap was initiated in October 2000 and
is scheduled for completion in September 2002.
The US
Department of Energy (DOE) launched the Generation IV initiative
in 2000. Today, it comprises 10 member countries plus Euratom,
with the goal of developing innovative nuclear systems (reactors
and fuel cycles) likely to reach technical maturity by 2030. The
ten countries are:
Argentina
,
Brazil
,
Canada
,
France
,
Japan
, the
Republic
of
Korea
, the
Republic
of
South Africa
,
Switzerland
, the
United Kingdom
and the
United States
.
Six nuclear
systems were selected;
·
Gas-Cooled
Fast Reactor
·
Lead-cooled
Fast Reactor
·
Molten
Salt Reactor
·
Sodium-Cooled
Fast Reactor
·
Supercritical-Water-Cooled
Reactor
·
Very
High Temperature Reactor
With the lofty
goal of achieving significant improvements in economic
competitiveness, nuclear safety, uranium resource economy and in
reducing long-life radioactive waste.
A technology
roadmap (http://gif.inel.gov/roadmap/),
initially prepared at the request of the US Congress, is now
used as a basis to structure and share the R&D effort among
the participating countries, in order to develop the selected
nuclear systems.
In the future,
other countries or international authorities could join the
Generation IV International Forum and the related R&D
effort.
Generation
IV – pooling international R&D
The founding
principle of the Generation IV International Forum (GIF) is its
members’ recognition of nuclear technologies’ role in
satisfying the world’s increasing energy needs – in the
context of sustainable energy development and climate change
prevention. This principle, laid down in the GIF charter,
expresses a common will to create a framework for international
R&D. This framework’s role is to define, develop and
enable the deployment of Generation IV nuclear systems. Most of
the research involved will develop within this multilateral
framework, over the coming decades.
What is
Generation IV?
Nuclear
systems can be classified according to the generation they
belong to. The US DOE distinguishes four generations:
·
Generation I was operational before the 1970s and
made use of natural uranium, to avoid the need for enrichment;
·
Generation II comprises light water reactors
deployed since the 1970s. They are still in operation;
·
Generation
III
involves optimizing the current reactors, in terms of economics
and safety. These reactors are likely to be deployed before
2010;
·
Generation IV comprises nuclear systems likely to
reach technical maturity by 2030. Their design will take
cognizance of the progress made in economics and safety. In
addition, the aim is for these reactors to support sustainable
energy development worldwide, and to open up the range of
nuclear systems’ applications to hydrogen generation for
transport (in addition to electricity production).
Generation
IV’s goals
International
consensus has been reached on the on the general goals and
criteria to be met by Generation IV nuclear systems. They will
have to be:
·
sustainable: the systems should offer efficiency
in the use of the natural resources, and should minimize
environmental impact (and, at the same time, minimize waste in
terms of mass, radio toxicity, residual power, etc.);
·
economically viable: economic considerations are:
the generating cost, which should be competitive when compared
with other energy sources; and the capital investment cost,
which should be low enough for the nuclear system under
development to remain accessible to a large number of countries
(total investment cost and specific investment cost ($/kWe)
refers);
·
safe and reliable: it is mandatory that future
reactors perform at least as well in terms of safety and
reliability as current reactors. In particular, key focus is to
be placed on eliminating, as far as possible, the need for
public evacuations from areas outside nuclear sites – in the
event of an accident, whatever its cause and extent of gravity;
·
Resistant to proliferation risks and likely to be
easily protected from external attack.
Besides
electricity generation, the Generation IV systems will offer
potential for the generation of hydrogen from water for use in
transport, seawater desalination, and heat generation for
industrial processes.
The
selection of Generation IV systems
From April
2001 to October 2002, technologies likely to meet the
above-mentioned criteria were identified by means of the
roadmap. The six selected concepts will be developed, in
multilateral co-operation, during a second phase of the
Forum’s activity.
The six
concepts, as presented in Figure 1, are:
·
GFR
: Gas-cooled Fast Reactor system cooled with helium;
·
LFR: Lead Fast Reactor cooled with lead or
lead-bismuth eutectic;
·
MSR: Molten Salt Reactor fuelled with molten
salts;
·
SFR: Sodium Fast Reactor;
·
SCWR: Super-Critical Water-cooled Reactor;
·
VHTR: Very High Temperature Reactor cooled with
helium at 1000°C at the core outlet, for efficient production
of hydrogen.

Among the six
selected systems, three have fast neutron spectra: the
GFR
, SFR and LFR. Two advanced gas-cooled systems (
GFR
and VHTR) rely on R&D which, to a large extent, follows the
same path. The SCWR was retained with a thermal neutron spectrum
as an intermediate step and a fast neutron spectrum as an
ultimate goal. The MSR stays in the running as a
non-conventional system. With the exception of the VHTR, all
these systems will be operated with a closed fuel cycle.
Key
comments relevant to the selection of the six Generation IV
systems:
·
Among the retained goals and criteria,
sustainability has been the most discriminating. This is
demonstrated by the majority of the systems having fast neutron
spectra and closed fuel cycles.
The grouping of systems in families – according to homogeneity
in performance and R&D needs – turned out to be important
for optimizing R&D efforts and structuring recommendations
along the lines of federal guidelines.
Selecting two gas-cooled systems (
GFR
and VHTR) was an acknowledgement of the interest in this coolant
for high temperature applications. A strongpoint governing this
selection is the consistency of the technology they employ,
enabling a sizeable common R&D pathway to be followed.
The issues which prevented the more innovative systems from
being selected concerned important uncertainties on: their
definition and performance, as well as on the prospects of
overcoming obstacles to their viability before the 2030
realization deadline.
Shared
R&D Efforts
Two main
phases of R&D were identified:
·
The ‘feasibility phase’ dedicated to resolving
technology showstoppers;
The ‘performance phase’, aiming at confirming and optimizing
the systems’ performance (which was the criterion emphasized
in the selection process).
The
feasibility studies for the more mature systems – i.e. the
Sodium Fast Reactor (SFR) and the Very High Temperature Reactor
(VHTR) – should last until 2008 and 2012 respectively; and the
performance studies, until 2015.
Regarding the
more futuristic systems – i.e. those having fast neutrons and
gas coolant (
GFR
), lead (LFR) and supercritical water (SCWR) – the feasibility
studies will continue through 2013 and 2015, and the studies of
optimization until 2020. The timeline for the molten salt system
will be longer.
While nuclear
R&D organizations will play an essential role in both the
feasibility and the performance studies, there will be
substantial participation from universities, other research
organizations and manufacturers. Countries that are not
currently GIF members will be able to join the R&D effort at
this stage.
Following the
feasibility and performance phases, it is expected that
international consortiums of manufacturers and R&D
laboratories will support demonstrations of the key
technologies. They will also identify which Generation IV
systems they are interested in launching commercially.
Progress in
preparing for the next phase of GIF
The GIF Policy
Group meeting, held in
Zurich
held on January 27 -27, 2004, led to progress on three main
topics: the co-operation agreement at system level, the
governance of the Forum, and relations with other organizations.
In
Zurich
,
the US DOE presented a draft system agreement – jointly
prepared by its State and Trade Departments – covering the
first 10 years of co-operation (and providing for extensions in
five-year increments). This project of agreement will establish
the R&D framework required to address Generation IV system
feasibility issues, as well as to confirm system performance,
established during the system selection process. Future phases
of demonstrating and commercializing the six selected nuclear
systems will be the subject of further agreements. The parties
to this system agreement are intended to be governmental
entities or mandated national laboratories. GIF members are to
be invited to give their input on the draft system agreement,
which is expected to be finalized by mid-2004
New
Reactor Designs
Overview
This issue paper
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 in the
U.S. Nuclear Regulatory Commission's certification or
pre-certification programs or 2) inclusion under the
international 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, are hyperlinked in the text.
Reactor vendors who put forward new designs anticipate that
their designs will meet commercial market needs including an
affordable, competitive construction cost and the usually low
operating costs of commercial nuclear reactors. Such views are
not assessed, though a section does identify public discussion
of efforts by 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 replicate
any of these designs for future construction. This is less
likely, however, because existing, operable reactors in the
United States were licensed during or before the 1970s.
Technology has progressed and any future construction should
incorporate more advanced designs that better meet today's
commercial and safety criteria.
There are possible
exceptions to the preceding statement. Three or four reactors in
the United States were partially built and still possess valid
construction licenses. These reactors are WNP-1 (Washington
State), Watt's Bar 2 (Tennessee), and Bellefonte 1 and 2
(Atlanta). Moreover, these construction licenses have recently
been extended to approximately the end of the present decade.
Construction on each unit was halted over a decade and a half
ago. Builders of these units, subject to the rules of their
licenses, have the right to resume construction on their
reactors, units that were designed during the 1970s or earlier.
Whether the construction will resume and whether former designs
will be continued remains to be determined. The owners of WNP-1
have announced plans to forgo their construction license to
allow for disassembly.
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 sent to a heat exchanger (called a
"steam generator" in PWRs) 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
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
units operating in the U.S. 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 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 also
possess some technology rights as well. Other major makers of
PWR reactors, including Framatome ANP and the Russian
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 coolant water directly into the steam (i.e. no heat
transfer) 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 though 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 since 1969 have been permanently shut down.
The average annual capacity factor for nuclear reactors in the
United States has been around 90 percent during the early
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.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.
Small prototypes now operate in China and Japan and additional
HTGR designs are promoted by firms in South Africa, the United
States, the Netherlands, and France. HTGRs use a gas- recently
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 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 their suitability for
hydrogen production. Advocates of HTGR designs hold that HTGR's
have high safety, low costs, and an ability to supply power to
smaller markets.
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 only very
limited market support, though "commercial" units
operate in Russia and France, and prototypes exist elsewhere,
notably Japan and India. "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 of natural uranium is the U-238 isotope that is not
directly usable as a reactor fuel. During the course of any
reactor operation a portion of the U-238 in the fuel is
converted to plutonium, primarily the useful Pu-239 isotope,
which provides much of the energy used in nuclear power
production. The bulk of the U-238 content in an LWR is not
converted to plutonium and the unconverted U-238 does not
contribute significantly to power generation. A breeder reactor
converts more U-238 to usable fuels than the reactor consumes.
Any unused fuel would have to be "reprocessed" before
some of the plutonium and the remaining U-238 would again be
usable. FBRs have, so far, proven to be more expensive to build
and operate than LWRs. It is not yet clear whether this is due
to the fact that most FBRs have been prototypes or if this
reflects underlying costs. The plutonium content of the
reprocessed fuel also raises concerns over weapons
proliferation. Many early FBR designs were prone to system
failures, though some, notably the BN-600 unit in Russia, have
operated over extended periods. Proponents of advanced reactor
designs believe that some commercial FBR designs could be
deployed prior to many other advanced, but untested commercial
designs.8
PHWRs have been
promoted primarily in Canada and India, with additional
commercial operating units in several other nations including
South Korea, China, Romania, and Argentina. Canadian designed
PHWRs are often called "CANDU" reactors.9
Siemens, ABB, 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 usually
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. Proliferation
issues related to the continuous process of refueling PHWRs have
raised some concerns as has spent fuel composition.10
The term gas-cooled
reactor (GCR) can be used ambiguously . HTGRs, for example, are
clearly a subset of GCRs that operate at higher temperatures. As
used here, GCRs are "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 have been built in France, Sweden, and
Japan. No GCR design has operated commercially in the United
States. Existing GCR designs have not been commercially
successful outside of the United Kingdom. Commercial GCRs11
in the United Kingdom have operated longer than commercial
reactors anywhere else in the world. Like the PHWRs, the
original GCR designs use natural uranium fuels, though newer
designs use slightly enriched fuels and are not confined to
uranium. 12
Other potential
designs for commercial reactors abound. They have not been
widely considered in recent history in the United States. There
is some experience with additional concepts elsewhere.
New
Designs
1.
Certified Designs
Following statutory
requirements, 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 common design
issues prior to construction. Design certification is an
optional process and may occur simultaneously with site
licensing.
Certification
Process for New Reactors in the United States, Design,
Type and Present State
| Reactor
Design |
Lead
Vender(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 |
Certification |
| ESBWR |
GE |
BWR |
Pre-certification |
| SWR-1000 |
Framatome
ANP |
BWR |
Pre-certification |
| ACR-700 |
AECL |
PHWR/PWR
hybrid |
Pre-certification |
| PBMR |
Eskom |
HTGR |
Pre-certification |
| GT-MHR |
General
Atomic |
HTGR |
Pre-certification |
| IRIS |
Westinghouse
BNFL |
PWR |
Pre-certification |
| EPR |
Framatome
ANP |
PWR |
No
application decision |
| ACR-1000 |
AECL |
PHWR |
No
application decision |
Note:
Reactor names are defined in the text.
Any new reactor
built in the United States over the next decade or so would
probably use designs either recently certified by the NRC or
that will be certified by the NRC in the near future. 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 US.
System 80+ (Westinghouse
BNFL): The System 80+ reactor is a PWR that was designed by
Combustion Engineering 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 that has been developed in Korea for
future deployment. I
ABWR
(General Electric, Toshiba, Hitachi): Of the three NRC-
certified ALWR designs only the ABWR has been deployed. Three
ABWRs operate in Japan, and four are under construction, two
each in Taiwan and Japan. While the ABWR design is usually
associated in the United States with General Electric, the
units now being built in Japan are products of Toshiba and
Hitachi. General Electric, Toshiba, and Hitachi have shown an
interest in building ABWRs in the U.S. There are many
variations in ABWR design. The most frequently mentioned
capacities in the 1250-1500 MWe range. Smaller and larger
designs exist depending on vendor. Vendors now claim costs for
the ABWR that have attracted some customer interest.
AP600
(Westinghouse BNFL): The AP600 is a PWR designed by
Westinghouse BNFL and certified by the NRC. The AP600, while
based on previous designs, has innovative passive safety
features that permit a greatly simplified reactor design.
Simplification has reduced plant components and construction
costs. The AP600 has been bid overseas but has never been
built. Westinghouse has recently de-emphasized the AP600 in
favor of the larger, though potentially less expensive AP1000
design.
The initial ALWR
reactors 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 still be a barrier to
commercial success in the U.S. The ABWR design however has many
variations and continues to be promoted in the U.S. by several
vendors. It is being considered for construction at Bellefonte
by the Tennessee Valley Authority (TVA).
2.
Undergoing Certification
Only one reactor
design is presently undergoing certification with the NRC,
although this situation could change shortly as additional
designs move from "pre-certification" to actual
"certification". The process of certification can take
several years and depends heavily on what design is proposed and
supported by potential vendors and buyers.
AP1000 (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 do the digits
mean the present capacity of the design. The most recent
AP1000 now has 1117 MWe capacity. The AP1000 is an enlargement
of the initial AP600, designed to increase the reactor's
target output by about 90 percent without significantly
increasing the total cost of building the reactor. Operating
costs are anticipated to be less than the AP600. While
Westinghouse BNFL owns rights to several other designs, the
AP1000 is the principal product that the company now promotes
in the United States. The AP1000 is a PWR with innovative,
passive safety features and a much simplified design that is
intended to cut the material and construction costs of the
plant.One consortium of 9 utilities called NuStart Energy is
promoting the AP1000 design.
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 manufacturers. Concepts such as the ESBWR, the
SWR-1000, and the ACR-700 appear to be much further along toward
potential deployment than, say, the IRIS and GT-MHR designs.13
Pre-certification, however, represents a vendor's intention to
proceed toward commercialization in the U.S. and perhaps
globally. Pre-certification is a less expensive stage of the
overall certification process. Actual certification procedures
are much more complex and many NRC costs are born by the
applicant.
ESBWR (General
Electric): The ESBWR14
is a new simplified BWR design being promoted by General
Electric. It constitutes an evolution and merging of several
earlier design ideas including the ABWR and other designs that
are no longer being actively pursued by GE. The intent of the
new design, which includes new passive safety features, is to
cut construction and operating costs significantly from the
ABWR design. GE is investing heavily in the ESBWR though the
design might not be available for deployment for several
years. The nine utility NuStart Energy group promotes the
ESBWR as well as the AP1000 design.
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 has also
recently begun SWR-1000 pre-certification with the NRC.
Literature on the design emphasizes the reactor's passive
safety features. Passive safety should also mean lower
construction costs though this is not emphasized by Framatome.
Information on the SWR1000 can be found on http://www.de.framatome-anp.com/anp/e/foa/anp/products/s112.htm.
The SWR-1000 presently has no US utility sponsor. Information
related to certification of the SWR-1000 can be found at http://www.nrc.gov/reactors/new-licensing/license-reviews/swr-1000.html.
ACR-700
(Atomic Energy of Canada Limited): AECL's "Advanced CANDU
Reactor" ACR-70015
has been developed over a lengthy period of time and is
considered an evolution from AECL's internationally successful
CANDU line of PHWRs. CANDU reactors have been more of a
commercial success than any other line of power reactors other
than the LWRs. One of the innovations in the ACR-700, 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-700 is a PHWR, a PWR, or a hybrid between the
two designs. This AECL has aggressively marketed the ACR-700
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 has
been shown by Dominion Resrouces regarding possible ACR-700
construction at North Anna (Virginia) and in Canada by
Canadian firms.
Pebble-bed
Modular Reactor (PBMR)
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. More recently the design
promoted and revised by the South African utility Eskom. Eskom
continues to partner in the design with BNFL among its'
investors. Recently Eskom is expected to receive approval to
build a prototype PBMR in South Africa. Certification
procedures in the U.S. have slowed but never has been
abandoned. At around 165 MWe the PBMR is one of the smallest
reactors now being proposed for the commercial market. This is
considered a marketing advantage because new small units
require less capital investments than larger new units. 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. Fuels used in the PBMR would
be more highly enriched uranium than is presently used in LWR
designs. The design is considered a possible contender for the
US Department of Energy's Next Generation Nuclear Plan (NGNP)
program in Idaho. Details regarding the PBMR design can be
found on http://www.pbmr.com/.
Information related to certification of the PBMR can be found
at http://www.nrc.gov/reactors/new-licensing/license-reviews/pbmr.html.
Gas-turbine
Modular Helium Reactor (GT-MHR)
(General Atomic): The GT-MHR is an HTGR design that has been
developed primarily by the U.S. firm, General Atomics. The
most advanced plans for GT-MHR development relate to building
reactors in Russia to assist in the "burn up" 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
below the 20 percent enrichment level that defines highly
enriched uranium. In initial designs, the conversion of the
energy in the heated helium coolant to electricity would be
directly in a gas turbine. There has been concern regarding
untested aspects of this technology. This has led some
potential sponsors to propose less innovative heat transfer
mechanisms to generate electricity. The U.S. utility, Entergy,
has participated in GT-MHR development and has used the name
"Freedom Reactor" for the design. Because coolant
temperatures arising from HTGR reactors are much higher than
from LWRs the design is viewed as a potential commercial heat
source. There has been particular attention to the design's
potential in a non-polluting method to produce hydrogen. The
GT-MHR is considered a potential contender for the US
Department of Energy's Next Generation Nuclear Plan (NGNP)
program in Idaho. 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 at
http://www.nrc.gov/reactors/new-licensing/license-reviews/gt-mhr.html.
International
Reactor Innovative & Secure
(IRIS) (Westinghouse BNFL): Westinghouse BNFL has promoted the
IRIS reactor design as a significant simplification and
innovation in PWR design. The reactor would be smaller than
most operating PWRs and would be much simplified. The IRIS
reactor includes features designs that are intended to avoid
loss of coolant accidents. Research has continued for some
time on the design and pre-certification is in process. The
IRIS is viewed as not ready for development during the present
decade, but may show potential during the next decade. IRIS
has a targeted 2015 completion date for the design. The design
presently has not utility sponsor but this is to anticipate
early commercial interest.Information on the IRIS can be found
on http://www.nei.org/index.asp?catnum=3&catid=712
and through http://www.nrc.gov/reactors/new-licensing/new-licensing-files/ml030780800.pdf.
Information related to certification of the IRIS can be found
at http://www.nrc.gov/reactors/new-licensing/license-reviews/iris.html.
4.
Anticipated for Possible Pre-Certification
Two designs, the
European Pressurized Water Reactor (EPR) and the ACR-1000,
have not been submitted for pre-certification in the United
States. Because of the attention that the designs are now
receiving, they are described below.
EPR
(Framatome ANP): Framatome ANP has not decided if it will
market its EPR in the United States. The EPR is a rather
conventional PWR unit though components have been simplified
and considerable emphasis is placed on reactor safety. The
design is being built in Finland. Additionally, the French
government has proposed building an EPR in France. Nuclear
power already constitutes over 75 percent of France's power
supply; therefore building a new reactor in France might
lead to decommissioning an existing reactor to make room in
the market for the base load power provided by an EPR.
France has proposed the EPR construction to China. The
proposed size of the reactor would be around 1600 MWe though
earlier designs were as large as 1750 MWe. Framatome has
indicated that it might seek US certification of the EPR
after European development proceeds.
ACR-1000
(Atomic Energy of Canada Limited): While AECL promotes its
ACR-700 design, an ACR-1000 is being designed as well. If
the scale economies attributed by Westinghouse BNFL to its
AP series and to GE's ABWR are valid, one might anticipate
parallel, cost-lowering results for the ACR series.
Advertised costs for the ACR-700 are already as low as any
design proposed for the near term. Promised construction
times of three years would set modern records for larger
reactors. Information on the ACR-1000 can be found on http://www.aecl.ca/index.asp.
5.
Generation IV Concepts
The U.S.
Department of Energy participates in the Generation IV
International Forum (GIF), an association of twelve nations
that seek to develop a new generation of commercial nuclear
reactor designs before 2030. Criteria for inclusion of a
reactor design for consideration by the 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)
4. Proliferation
resistance (does not add unduly to unsecured nuclear material)
and
physical protection; (secure from terrorist attacks)
During 2002, GIF
members agreed to concentrate their efforts and funds on six
concept designs that they seek to become commercially viable
between 2015 and 2025. There is thus some leeway between the
2030 target for the GIF program and the targets for individual
concepts. Individual GIF participant nations are free to
pursue any technology they chose. 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), has 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://www.nuclear.gov/
and describes each design in some detail including reactor
schematics. Each design is evolutionary; thus while the
following descriptions involve comparisons, these analogies
should be interpreted with caution. Gen IV programs are
summarized on http://www.inel.gov/initiatives/generation.shtml.
Gas-cooled
Fast Reactor
(GFR): The GFR uses helium coolant directly 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 is 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.
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 been difficult to handle.The LFR
uses molten lead or a lead-bismuth alloy as its coolant. One
design favored under the Generation IV would result in long
periods between refuelings, 15-20 years. Similar designs
have been investigated in Russia. 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.
Molten
Salt Reactor
(MSR): The MSR involves a circulating liquid of sodium,
zirconium, and uranium fluorides as a reactor fuel. 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. The design could use a wide variety of
fuel cycles. Temperatures for electricity production 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 never were implemented for
commercial uses. During 2003, the MSR was down rated within
the Gen IV program because it was seen as too distant into
the future for inclusion within the Gen IV schedule.
Sodium-cooled
Fast Reactor (SFR):
Sodium-cooled fast reactors have been the most popular
design for breeder reactors. Designs have been proposed
under the "Technological Roadmap" ranging from 150
to 1700 MWe. 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 reflects the considerable research that the
design has already received. This date seems to be lagging
as the VHTR gains favor. Earlier prototypes have already
been built in France, Japan, Germany, the United Kingdom,
Russia, and the United States since 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
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 the
BWR and PWR. SCWRs would operate at higher temperatures and
thermal efficiencies than present LWRs. The reference plant
would be 1700 MWe, toward the upper end of present LWR
designs. The deployment target date is 2025. Some GIF
participants favor the design. Most research on the design
has been in Japan. Designers intend the SCWR to be much less
expensive to build than today's LWR units though some of the
economies appear to be shared by units now undergoing
certification. Operating cost savings are also anticipated.
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. 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 would
provide process heat that could be used in industrial
activities including hydrogen production and desalinization.
Electricity generation targets have not yet been set.
Deployment is targeted for 2020, earlier than most other
Generation IV designs. The VHTR is now the favored design in
the US, where it is the basis for the Next Generation
Nuclear Plant (NGNP) program in Idaho. France also favors
the design.
Each GIF project
involves new or untested concepts in reactor design. It would
be surprising if every design concept met the program's
initial targets. The research involved in the program has the
potential to contribute to the understanding of alternative
types of commercial nuclear power and process heat even if
individual projects do not meet 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 technology has been too
expensive to compete in the commercial marketplace. There have
been relatively few orders for new nuclear power plants during
the last two decades, not just in the United States and
Canada, but also in Western Europe. Interest in new nuclear
power units has recently focused on Asia and to a lesser
extent in Eastern Europe. New orders in Finland and
potentially France follow a long period of market inactivity.
Reactor vendors have not ignored the message that their
product has recently involved high construction costs and long
construction periods. Vendors are attempting 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 their
designs, though such prices are often confined to those parts
of construction that the vendors actually control, the
"nuclear island" that they designed.
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 by the Federal Energy
Regulatory Commission (FERC), have been roughly the same as
and 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 gas-fired
generation units. 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.
Discussions of estimates of the capital and operating cost of
new power generation units can be found on in the "Issues
in Focus" section of the Annual
Energy Outlook and in the Electricity Section of the
Assumptions for the Annual
Energy Outlook.
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),
website: http://www.iaea.org/worldatom
- "A
Roadmap to Deploy New Nuclear Power Plants in the United
States by 2010," United States Department of Energy
Office of Nuclear Energy, Science and Technology and its
Nuclear Energy Research Advisory Committee Subcommittee on
Generation IV Technology Planning (October 31, 2001),
website: http://www.nuclear.gov/
- Scully
Capital, "Final Draft, Business Case for Nuclear
Power Plants, Bringing Public and Private Resources
Together for Nuclear Energy" (July 2002) (available
through United States Department of Energy
Office of Nuclear Energy, Science and Technology, website:
http://www.nuclear.gov/
- "A
Technology Roadmap for Generation IV Nuclear Energy
Systems (GIF-002-00)" U.S. DOE Nuclear Energy
Research Advisory Committee and Generation IV
International Forum (December 2002), website: http://www.nuclear.gov/
Summary
and Potential
There are early
signs that the nature of the nuclear reactor market might be
changing. Finland is building based on Framatome ANP's EPR
design for a "fifth nuclear reactor". France also
plans to build an EPR. More recently Bulgaria has discussed
building a new nuclear reactor at Belene. Belene was
originally to be a Russian-designed VVER-1000 unit using
equipment already owned by and located in Bulgaria.
Competitive submissions from several vendors for alternative,
newer designs have been considered. It is unclear if the
original design plan for Belene will move forward or if new
designs will be slated.
The United States
is funding a program called Nuclear Power 2010 that seeks to
build at least two nuclear power reactors by the mid 2010's.
Supporting this has been proposed Federal energy legislation.
Meeting the target would be a challenging task and the
proposed legislation is still being debated .
1A
large number of reactor designs have been excluded from the
discussion. These include reactors promoted overseas by
nations such as Russia, India, Argentina, Korea, Canada, and
China, as well as numerous smaller or even portable reactors
that are being examined worldwide, including in the United
States. Also excluded is the International Atomic Energy
Agency's International Project on Innovative Nuclear Reactors
and Fuel Programs (INPRO) that covers territory similar to the
GIF program in addition to other promising designs. GIF
designs have been more heavily promoted within the United
States.
2The
one that is not operational, Brown's Ferry 1, has been shut
down since 1985, but has not given up its operating license.
The plant's owner-operator, the Tennessee Valley Authority,
intends to restart the reactor in 2007.
3The
terms "cooled" and "moderated" are
important because they define reactor categories. Cooling in a
reactor refers to the process and medium by which heat is
transferred from the reactor core to the steam supply cycle of
the nuclear power plant. "Moderating" is a concept
unique to nuclear power. A moderator controls the rate of the
nuclear power reaction and thus the amount of heat that is
generated. In a light water reactor ordinary water serves both
functions. Light water contains the same isotopes of hydrogen
and oxygen as naturally occurring water. Heavy water contains
a different, heavier isotope of hydrogen known as deuterium.
Beyond the point that these conditions define reactor types,
this will not matter in the discussion of existing reactors.
It does matter for the group that will be discussed under
"Generation 4" reactors.
4Exceptions
include Canada, the United Kingdom, India, and part of
Russia's industry.
5
Prior to 1969, some smaller commercial reactors were placed in
service. All have been retired.
6
This is based on Utility Data Institute/Resource Data
Internationl Compilations of FERC Form 1 data.
7The
discussion here does not directly address
"enrichment" the process by which the U-235 content
of nuclear fuel is increased.
8This
latter statement is based on "A Technology Roadmap for
Generation IV Nuclear Energy Systems".
9Candu
is a contraction of the term "Canadium deuterium".
Canada has an interesting and unique nuclear power history
which is covered by the book, Atomic Energy of Canada
Limited, Canada Enters the Nuclear Age.
10Inspectors
of nuclear power plants have a preference for plants such as
the LWRs that are refueled in batches rather than the
continuous fueling of PHWRs. Batch refueling is more easily
monitored and occurs at intervals of one to two years.
11
Not the AGRs.
12
Most designs of PHWRs also use natural uranium fuels. However,
variations in fuel type are possible at any PHWR with
plutonium and thorium fuel content subject to particular
interest and experimentation.
13
This sentence is a good example of the acronyms that overwhelm
the nuclear steam supply system (NSSS) industry. Several of
these acronyms no longer have any meaning in "words"
while others have only limited actual meaning. They are
defined below when possible.
14
The term ESBWR is now called the "EconomicSimplified
Boiling Water Reactor". Definitions of the initials have
changed overtime.
15
ACR is usually read to mean "Advanced CANDU
Reactor".
The
Principles of Nuclear Power
- In naturally
occurring uranium, 0.7% of uranium is of a particular type
(isotope) of uranium (U235) which spontaneously splits
(fissile material) to emit a tiny particle (a neutron). If
this neutron hits another U235 atom, it too will split (a
fission) to produce two more neutrons (chain reaction).
- If the
concentration of U235 is sufficient (a critical mass), the
process will be self-sustaining (the plant is `critical'),
producing large quantities of heat in the `core' of the
reactor.
-
- Two important
ingredients are needed to control the process and to
utilise the heat, the moderator and the coolant. A
moderator is a substance which neutrons collide with but
`bounce off' without absorbing too much energy and without
itself being split. It controls the amount of neutrons
escaping from the core before they have hit another U235
atom. A good moderator is one which absorbs the least
energy and does not absorb the neutrons before they split
another uranium atom. Graphite is an excellent moderator;
ordinary water is a poorer moderator but is much cheaper.
If water is used, the U235 content must be increased
(enrichment) to about 3 per cent to allow a chain reaction
to take place. A rare isotope of hydrogen (deuterium) can
be used to make so-called heavy water (deuterium is twice
the weight of normal hydrogen) and this is also an
excellent moderator.
-
- In so-called
fast (breeder) reactors (as opposed to the thermal
reactors described above), no moderator is used and some
of the neutrons escape the core and strike a `jacket' of
uranium where they convert the unused part of the uranium,
U238, to fissile material, plutonium, which can be used as
a reactor fuel. The jacket is processed to isolate the
plutonium for use in more fast reactors. The attraction of
this design is obvious, it can use almost 100 per cent of
naturally occurring uranium instead of the 0.7 per cent
thermal reactors achieve. The disadvantage is equally
obvious: it requires the separation, transport and
widespread use of the material used to make nearly all
nuclear weapons and is regarded as a serious proliferation
risk. The technical attractions of the design have lead to
huge amounts of public money being spent on this
technology. However, in practice, all prototype plants
have proved most unreliable and the technology is now all