LEAD-COOLED FAST REACTOR
Lehrstuhl für Nukleartechnik - Technische Universität München
Boltzmannstr. 15 85747 Garching
www.ntech.mw.tum.de
1
1 LEAD-COOLED FAST REACTORS
1.1 CONCEPT DEVELOPMENT
The Generation IV Roadmap selected the lead fast reactor (LFR) concept as one of
the six technologies for further development under Generation IV. GIF has
established a Steering Committee for the development of LFR, with participation of
Euratom, Japan, the Republic of Korea and the United States. There are active and
large development programs in Europe and Russia to develop project for the
realization of Lead, individuated as the reference coolant option, and Lead-Bismuth,
individuated as a backup option, systems.
Sixteen European Organizations are collaborating together with two additional
institutions from US and Korea to present to the European Commission the proposal
for a Specific Targeted Research and Training Project (STREP). This project is
devoted to the development of a European Lead-cooled System (ELSY). This system
will constitute the larger lead-cooled reactor of GEN IV. ELSY aims to demonstrate
the feasibility to design a competitive and safe fast power reactor which complies
with al the GEN IV goals and gives assurance of investment protection.
Russia joining GEN IV it is expected will strongly reinforce the international effort
towards LFR development. In fact, Russia gained a unique practical experience in
operation of reactors with lead-bismuth coolant that in total reaches the amount of
80 reactor-years. Eight reactors were installed in submarines of the former USSR
Navy. Besides, two full scale reactor prototypes were constructed and operated in
Obninsk and Sosnovy Bor. Further, a Pb-Bi loop were tested at the Kurchatov
Institute. In the last 15 years Russian institutions inside the Federal Agency for the
Atomic Energy and the Russian Research Center “Kurchatov Institute” carried out
large R&D work in order to demonstrate that heavy metal cooled fast reactors can be
developed in a limited period of time with the wide use of established technologies
and side by side with traditional sodium-cooled fast reactors.
One of the principal purposes of the GIF R&D plan is to identify the priorities for
common and coordinated LFR research. The purpose is to pursue a dual-track
approach leading to the development of a small transportable system and a
moderate- or large-scale power plant. The major topics are the system design, fuel
development, lead technology and materials, component development, balance of
2
plant, the hydrogen production and the demonstration. The GIF Expert Group
recognizes two major thrusts in the LFR program and they are:
The Small Secure Transportable Autonomous Reactor (SSTAR);
The European Lead-cooled System (ELSY)
Other activities have been planned and they will lead to the development of
advanced materials for Lead Bismuth Eutectic (LBE) applications.
Priorities in each category are individuated based on the dual-track approach
explained in Figure 1.1.
Figure 1.1: LFR R&D conceptual framework and schedule
(source “Coolants and Innovative Reactor Technologies”
AIX EN PROVINCE - 27th November, 2006, L. Cinotti)
The institutions involved in the ELSY project are listed in Table 1.1.
3
Table 1.1: Institutions involved in the ELSY Project
It is possible to consult a complete database of information concerning Liquid Metal
Cooled Fast Reactors at the following internet address:
http://www.iaea.org/inisnkm/nkm/aws/frdb/index.html.
4
1.2 TECHNICAL ASPECTS
Different concepts for the design of a lead cooled fast reactor are under study and
they are divided essentially in two different groups:
a small transportable system
a moderate- or large-scale power plant
The first reactor type is named Small Secure Transportable Autonomous Reactor
(STARR) will be a small, modular, fast reactor. The main mission of the 20 MWe
(45MWth) reactor is to provide incremental energy generation to match the needs of
developed countries and remote communities without electrical connections. This
will be a niche market product where costs higher than those of a large-scale nuclear
power plant remain competitive.
The second reactor type is the European Lead-Cooled System (ELSY). This project
aims to demonstrate that it is possible to design a competitive and safe lead fast
power reactor using simple engineered features. It will be a large-scale power
reactor that will be economically productive on the existing well interconnected
grids.
1.2.1 Small Secure Transportable Autonomous Reactor (STARR)
The Small Secure Transportable Autonomous Reactor (SSTAR) (Figure 1.2) is a 20
MWe (45 MWth) exportable, small, proliferation-resistant, fissile self-sufficient,
autonomous load following, and passively safe lead-cooled fast reactor (LFR)
concept for deployment at remote sites. SSTAR is a pool-type reactor and is
currently at a pre-conceptual level of development.
Potential customers for SSTAR include: clients looking for energy security at small
capital outlay; cities in developing nations; and deregulated power producers in
developed nations. SSTAR makes extensive use of inherent safety features; most
notably, natural circulation heat transport, Pb coolant, and transuranic nitride fuel.
The SSTAR nuclear power plant incorporates a supercritical carbon dioxide (S-CO2)
Brayton cycle power converter for higher plant efficiency and lower balance of plant
costs. The efficiency of the S-CO2 Brayton cycle increases as the reactor core outlet
temperature increases; an efficiency of about 44% can be attained for a turbine inlet
temperature of about 550 °C. To take advantage of the economic benefits of such a
5
high plant efficiency, there has been interest in operating at higher Pb coolant
temperatures. In particular, a peak cladding inner surface temperature of 650 °C has
been an objective.
Figure 1.2: SSTAR Modular Lead-Cooled Fast Reactor
SSTAR is scalable to a higher power level of 181 MWe (400 MWth); this is the
STARLM (Secure Transportable Autonomous Reactor with Liquid Metal) concept.
STAR-LM is a scaled-up version of SSTAR for high efficiency electric power
production with optional production of desalinated water using a portion of the reject
heat. The STARLM reactor vessel size (16.9 m height by 5.5 m diameter) is assumed
to be limited in height by a rail shipment length limitation of 18.9 m. The power level
of 400 MWth approaches the maximum value at which heat transport can be
accomplished through single-phase natural circulation given the reactor vessel
height limitation. The scaled-up version can alternately be used for hydrogen and
oxygen generation using a Ca-Br thermochemical (“water cracking”) cycle, if
cladding and structural materials for operation with Pb up to about 800 °C can be
developed; this high temperature version is named STAR-H2.
Notable achievements of SSTAR development include:
Pb coolant;
30-year core lifetime;
Average (peak) discharge burnup of 81 (131) MWd/Kg of Heavy Metal;
Burnup reactivity swing < 1 $;
6
Peak cladding temperature = 650 °C;
Core outlet/inlet temperatures = 564/420 °C;
Peak transuranic nitride fuel temperature = 882 °C;
Small shippable reactor vessel (12 m height by 3.23 m diameter)
Autonomous load following;
Supercritical CO2 Brayton cycle energy conversion efficiency = 44.2 %;
Plant efficiency = 43.8 %;
Cost of energy generation < 5.5 $-cents/kWh (55 $/MWh).
Conditions, dimensions, and other parameters for SSTAR are included in Table 1.2
and Table 1.3.
Table 1.2: Conditions and dimension for SSTAR (1/2)
Table 1.3: Conditions and dimension for SSTAR (2/2)
7
1.2.2 Lead-Cooled System (ELSY)
The ELSY power plant (Figure 1.3) is tentatively sized at 600 MWe because only
plants of the order of several hundred MWe are expected to be economically
affordable on the existing, well-interconnected grids of Europe. The mass of lead of a
LFR is worldwide considered a critical issue for the reactor vessel which can limit the
plant power. For this reason a preliminary mechanical verification, including seismic
loads, has been performed from the beginning of the design activity based on
preliminary parameters. The activity is aimed to confirm that the relatively small
vessel dimensions are realistic also thanks to innovative solutions of the primary
system layout. A LFR of a power larger than a medium power is potentially feasible
according to these preliminary evaluations.
The main reasons for selecting lead as primary coolant for ELSY are:
lead is much more abundant (and less expensive) than bismuth (pure lead as
coolant offers then enhanced sustainability)
the use of lead strongly reduces the production of the highly radioactive
decay-heat generating polonium in the coolant with respect to LBE
8
operation at a higher lower limit of the thermal cycle, required by the use of
pure lead, would be necessary also in the case of LBE to improve plant
efficiency and to avoid the excessive embrittlement of structural material
subjected to fast neutron flux
Figure 1.3: Preliminary scheme of the ELSY Reactor
9
Tentative parameters for ELSY are included in Table 1.4.
Table 1.4: Tentative parameters of the ELSY plant
1.2.3 RBEC Lead-Bismuth fast reactor-breeder
Another important reactor concept is under development in Russia. It is the RBEC
lead-bismuth fast reactor-breeder. It is oriented at the deployment in the near future
and based on established decisions and technologies on the use of lead-bismuth
coolant experienced in nuclear submarines; layout, fuel type, steam parameters are
close to those used in existing reactors. The aim of the RBEC project is the creation
of a nuclear steam-generating power plant on the basis of Russian experience in
design and operation of fast reactors and liquid-metal technology.
High self-protection level should be provided by
inherent core safety properties
thermal-physical properties of lead-bismuth coolant
use of natural circulation for emergency core cooling
application of passive safety systems along with traditional active ones
10
qualitative factory fabrication of the equipment
The three-circuit scheme was implemented in the reactor design of 900 MWth and
340 MWe power unit developed by Russian organizations - OKB Gidropress, RRC
Kurchatov Institute, and IPPE. The design and thermal-hydraulic parameters of
RBEC are based, as much as possible, on technical decisions proved in BN-type
reactors cooled by sodium, and they correspond to existing experience on fuel,
structural materials, and technology of liquid-metal coolant.
The RBEC reactor facility (Figure 1.4) contains the following main systems:
primary system structurally made as a monoblock unit
intermediate (secondary) system
turbine system
air emergency core cooling system
refueling system
system for gas heating or emergency cooling of monoblock vessel
system for electric heating of secondary circuit
system for filling and drainage of primary and secondary coolant
clad failure detection system
system of the primary and secondary coolant technology
control and protection system, automatic control, etc.
11
Figure 1.4: General view of RBEC reactor
12
The RBEC major characteristics are given in Table 1.5.
Table 1.5: RBEC Reactor major characteristics
Two types of MOX fuel with different Pu content are used in fuel rods to flatten the
power density radial distribution. The central low-content zone consists of 121 fuel
assemblies with 27.5% Pu content in fuel rods. The high-content zone includes 132
fuel assembly with 37.1% Pu content in fuel rods. The core is surrounded by 126
assemblies of radial blanket with fertile rods of depleted uranium carbide. 192
assemblies of neutron reflector are installed around the core.
13
1.3 TECHNICAL PROBLEMS
Basic properties of the considered coolants together with lead/bismuth eutectic are
summarized in Table 1.6.
Table 1.6: Basic physical properties of liquid metal coolants
The choice of lead and lead-alloys as coolants is motivated:
by their high boiling temperatures, which avoids the risk of coolant boiling,
by the fact that lead and lead-alloys are compatible with air, steam, CO2, and
water, and, thus, no intermediate coolant loop is needed as in the sodium-
cooled system,
Lead-bismuth eutectic provides a low melting point (398 K) limiting problems
with freezing in the system and features a low chemical activity with water
and air excluding the possibility for fire or explosions.
A drawback connected with lead/bismuth is the accumulated radioactivity in
lead/bismuth (mainly due to the á-emitter 210Po, T1/2 = 138 days), which could pose
difficulties during fuel reloading or repair work on the primary circuit. Using only Lead
the production of highly radioactive, and hence decay heat generating polonium is
much lower than in the case of LBE.
The omission of bismuth in the coolant reduces therefore problems associated with
decay heat removal. However, Fomichenko reports that IPPE Obninsk staff has
developed methods to cope with the polonium during refueling and maintenance.
Lead is considered as a more attractive coolant option than lead/bismuth mainly due
to its higher availability, lower price and lower amount of induced polonium activity
(by a factor of 104), as given in a publication about BREST- 300 LFR reactor design
like Fomichenko reports. Pure lead has a melting temperature of 601 K, which
narrows in the reactor’s operational interval to about 680-870 K. However, after more
research, higher outlet temperatures will eventually be possible. Redundant electrical
Notes:
14
heaters are proposed to be introduced in order to avoid problems with freezing and
blockages in fresh cores. Operation at higher temperature, required by the use of
pure lead, would generally be necessary also in the case of LBE to improve plant
efficiency and to avoid excessive embrittlement of structural material submitted to
fast neutron flux at low-temperature.
Positive and negative qualities of lead-based coolant for fast reactors have been
individuated by Fomichenko and by Cinotti respectively in their works and have been
summarized in the following table (Table 1.7).
Table 1.7: Positive and negative qualities of lead-based coolant for fast reactors
Properties Positive features Negative features
Low chemical activity
there is no fire and explosion danger, when coolant
contacts air or water;
there is no loss of coolant probability caused by
coolant burning out;
reactor design can be significantly simplified (for
example, two-circuit scheme can be used, steam
generator design can be simplified, etc.).
High corrosion activity
special on-line control of oxygen concentration in
primary coolant is required to form and maintain
protection oxide films on fuel cladding surfaces and
simultaneously to prevent precipitation of solid
oxides in cold sites of the primary circuit (heat
exchanger);
special measures are required to remove corrosion
products from the coolant;
positive reactivity effect is caused by solution of core
structural materials in the coolant and their transfer
from the core;
coolant temperatures and velocities are limited.
High boiling point
high temperatures and, hence, high efficiency can be
provided at low primary pressure without coolant
phase change; low primary pressure enhances
reactor safety and reliability, allows to simplify
reactor design and facilitate fuel rod operational
conditions;
there is no probability of cladding overheating
15
caused by departure from nucleate boiling;
there is no loss of coolant probability caused by
coolant boiling out;
there is no probability of overpressure shocks on
reactor equipment caused by coolant phase
changes;
positive reactivity can be inserted if coolant
overheating causes melting and flowing up of
structural materials out of the core.
Thermal phys. properties
high enough heat capacity of reactor circuit
eliminates fast temperature increases in accidents;
lower specific heat capacities and thermal
conductivity compared to sodium coolant.
High melting point
loss of coolant after circuit break is limited because
of fast coolant freezing and possible closing of the
break;
rapid freezing of coolant eliminates deep penetration
of radioactive coolant in the environment after
accident with primary circuit break;
high temperatures in all circuits should be provided
by special electrical heating system during reactor
startup, repair and maintenance, and shutdown;
special measures are required to exclude probability
of coolant freezing in heat exchangers in operational
and transient conditions.
High density
probability of secondary critical mass formation after
core degradation is low because coolant density is
close to or higher than fuel density and coolant flow
can distribute fuel fragments over primary circuit;
probability of vapor or gas entrainment in the core is
low due to high coolant buoyant force;
powerful and reliable pumps are necessary;
high requirements should be met to seismic stability
of the facility;
reactor vessel and support structures should have
high strength;
special measures should be envisaged to eliminate
16
flowing up of fuel assemblies caused by high coolant
buoyant force;
high erosion of structural materials limits coolant
temperatures and velocities; erosion of core
structural materials and their removal from the core
cause insertion of positive reactivity.
Nuclear parameters
low inelastic scattering and radiative capture cross
sections are favorable for improving breeding and
reducing void effect;
high elastic scattering cross sections are favorable
for albedo parameters of neutron reflectors,
surrounding the core and for reduction of activity
accumulation in coolant and reactor structures;
lead-based coolants themselves are very good matrix
for final disposal of radioactive products accumulated
in coolant;
high radioactivity in operation due to polonium
formation;
rather high long-lived activation which eliminates or
significantly restricts re-utilization of the coolant.
Toxicity lead is toxic.
The choice of a large reactor power suggests the use of forced circulation to shorten
the reactor vessel avoiding excessive coolant mass and alleviating mechanical loads
on the reactor vessel. The needed pump head, in spite of the higher density of lead,
can therefore be kept low (of the order of one to two bars) with reduced requirement
of pumping power. In fact, thanks to the favorable neutronic characteristics of the
coolant, the fuel rods of a lead-based coolant reactor, similarly to LWRs, can be
spaced further apart than in the case of sodium as a coolant and this will result in
low pressure drop through the core.
Therefore a simple gas lift as pumping system could be selected, instead of
mechanical pumps, to enhance the primary coolant natural circulation to the
specified flow rate. A test section of this gas lift system has been installed in the
CIRCE facility (at the ENEA site of Brasimone in Italy) with one full-scale riser pipe.
The test result confirms the suitability of gas lift for a small-power reactor, but shows
also decreasing efficiency at the higher flow rates, a fact that makes its applicability
17
questionable for a large plant such as ELSY for which, therefore, it is envisaged the
use of mechanical pumps.
Regarding the fuel, MOX (Mixed Oxide) fuel is considered like the reference for the
short term deployment in order to not introduce additional risk of delay for
deployment. On the long run, clear priorities have been not yet defined but the ELSY
project will at least provide two type of information:
the incentive to develop nitride fuels,
the capability of the system to accept fuel containing MA (Minor Actinides).
For the moment because of limitation on the materials database, it is not possible to
use the high potential offered by the high temperatures reachable by the use of lead
or lead-based coolants, in order to increase the efficiency of the system for the
energy and hydrogen production. This will require the development of new materials
both for mechanical components and fuel cladding.
18
1.4 ECONOMIC ASPECTS
Regarding economics, fast reactors were earlier considered more expensive to build
and their electricity generation cost higher than that of LWRs. However, Tucek at al.
in his work reports that in the last few years several Russian publications have
indicated that the lead/bismuth-cooled SVBR-75/100 is cheaper to build than all
other reactor types and that the electricity generation cost is even lower than that of
gas-fired plants, see Table 1.8.
Table 1.8: Economic comparison of LFR, SFR and Gas-fired plant
The reasons for this are that no intermediate coolant loop is needed for an LFR, and
less safety-related systems have to be built. The prolonged, 8-year fuel cycle is
helping to get the electricity generation cost down, too. Note that in a true LFR, lead
would be used instead of the lead/bismuth. Since lead is about 10 times cheaper
than lead/bismuth, the capital cost for LFR may be even lower than envisioned for
SVBR-75/100.
Lead-cooled fast reactor (LFR) concepts address the following objectives:
the ability to produce hydrogen,
smaller distributed plants,
making the plants increasingly environmentally benign,
the ability to load-follow to match production with need.
The last three considerations (small plants, reduced environmental footprint, and
load following) can be also achieved using other fast reactor concepts (sodium, salt,
or gas cooling). Because of the higher boiling temperature of lead and lead-bismuth,
lead-alloy systems are better positioned to link to high-temperature hydrogen
production than the sodium-cooled systems.
19
Cinotti et al. in their work present the economic issues of Lead Fast Reactors linked
to technical and safety issues. He reports that the cost advantage features of the
LFR must include
low capital cost,
short construction duration,
low fuel and low production cost.
The economic utilization of MOX fuel in a fast spectrum has been already
demonstrated in the case of the SFR, and no significantly different conclusion can be
expected for the LFR except from improvement due to the harder spectrum.
Because of the favorable characteristics of molten lead, it will be possible to
significantly simplify the LFR systems in comparison with the well known designs of
the SFRs, and hence to reduce its overnight capital cost, which is a major cost factor
for the competitive generation of nuclear electricity. A simple plant will be the basis
for reduced capital and operating cost. A pool-type, low-pressure primary system
configuration offers great potential for plant simplification. The use of in-vessel
Steam Generator Units (SGU’s) and the consequent elimination of the intermediate
circuit, typical of sodium technology, are expected to provide competitive generation
of electricity in the LFR.
This approach is possible because of the absence of fast chemical reactions
between lead and water, although the steam generator (SG) tube rupture accident
(i.e., pressure waves inside the SGU) must be considered in the design. The
configuration of the reactor internals will be as simple as possible. The very low
vapor pressure of molten lead should allow relaxation of the otherwise stringent
requirements of gas-tightness of the reactor head and possibly allow the adoption of
simple fuel handling systems. Corrosion by molten lead of candidate structural steels
for the primary system will be minimized by limiting the core outlet temperature.
Considering that there will be no intermediate circuit to degrade the thermal cycle
and that the expected core inlet temperature of about 400°C is relatively high, the
adoption of a high-efficiency water-steam supercritical cycle is possible. Additionally,
a supercritical carbon dioxide Brayton cycle energy conversion system can be
considered.
20
For small, transportable systems, a limitation to the risk to capital results from the
small reactor size. In addition, and with particular relevance to the moderator large-
size central station system, a reduction in the risk to capital results from the potential
for removable/replaceable in-vessel components.
The Idaho National Laboratory (INL) reports in its FY2005 Ten years Program Plan
that Overnight and generation costs remain to be estimated and that Financial risk
remains to be quantified.
Objectives of future works at the INL are to demonstrate the viability of reducing
costs by taking advantage of LFR system attributes that enable savings such as
system simplification through
elimination of the need for an intermediate heat transport circuit;
elimination of main coolant pumps;
autonomous load following that simplifies the control system and reduces
operator requirements
utilization of S-CO2 Brayton cycle power conversion that offers higher plant
efficiency together with smaller, simpler, and fewer balance of plant
components
small plant footprint
factory fabrication that reduces component costs
modular transport and installation at the site that reduces construction time
and costs.
The small modular plant requires a smaller outlay of funds and provides a shorter
construction time. When the plant goes online, it becomes a source of positive
cash flow that can be applied to financing the construction of the next module
and so on. An objective of future works at the INL is also to establish the viability
of this approach. An economic objective of passive safety is to demonstrate the
viability of minimizing the threat to investment in the plant due to postulated
accidents or sabotage.
21
1.5 ENVIRONMENTAL AND SOCIAL ASPECTS
Cinotti at al. in their work delineates a LFR system like an unattractive route for
diversion of weapon-usable material. In fact the use of a MOX fuel containing MA
increases proliferation resistance. The use of a coolant chemically compatible with
air and water and operating at ambient pressure enhances Physical Protection.
There is reduced need for robust protection against the risk of catastrophic events,
initiated by acts of sabotage because there is a little risk of fire propagation and
because of the passive safety functions. There are no credible scenarios of
significant containment pressurization.
Under the point of view of safety, lead is an inherent safe coolant because even in
case of leakage the coolant would solidify without significant chemical reactions
affecting the operation or performance or surrounding equipment or structures.
Also Fomichenko in his work reports that loss of coolant accident in reactors with
lead-based coolant is not accompanied by significant radioactivity release into the
environment because this accident and activity release are limited by rapid freezing
of the coolant and formation of protective oxide layer on its surface.
In case of lead as coolant, the system could reach a higher grade of safety. A severe
re-criticality is prevented because lead has a higher density than those of the oxide
fuel or of the low density metal fuel, and its natural convection flow will prevent fuel
aggregation with the possible formation of a secondary critical mass.
Under the point of view of sustainability, lead will allow a better resource utilization.
Lead presents very low neutron absorption and moderation, it makes possible an
efficient utilization of excess neutrons and reduction of specific uranium
consumption. Reactor designs can readily achieve a breeding ratio of about 1, and
long core life and a high fuel burnup can be achieved. Furthermore also the
production of wastes will be minimized and their management will be easier. A fast
neutron flux significantly reduces waste generation, Pu recycling in a closed cycle
being the condition recognized by GEN IV for waste minimization. The capability of
the LFR systems to safely burn recycled minor actinides within the fuel will add to
the attractiveness of the LFR.
22
Fomichenko in his work report a comparison of the long-lived residual radioactivity
for sodium, lead-bismuth and natural lead. The main conclusions of the analysis
were the following:
Sodium: in case of operation without fuel rod failures, after-irradiation cooling
of sodium for 50-60 years would be enough to use sodium again in any way
or to return it in the environment
Lead-Bismuth: because of high activity of bismuth, this coolant should be
classified and treated as radioactive waste practically forever if special
technology is not developed for slearing the activity;
Lead: situation for lead coolant is not so unequivocal as for lead-bismuth
coolant, but its repeated utilization only in radioactive-dangerous
technologies or final disposal are the most possible decisions for natural lead
coolant.
Fomichenko then lists a possible solution to the high activity of natural lead. He
suggests the use as a fast reactor coolant isotopically pure 208Pb which has excellent
nuclear properties and does not lead to formation of long-lived radionuclides during
irradiation. He reports also that lead and lead-based coolants have are a very good
matrix for final disposal of radioactive products accumulated in coolant.
23
1.6 BIBLIOGRAPHY
LFR “Lead Fast Reactor” L. Cinotti (Ansaldo Nucleare), C. Fazio (FZK), J. Knebel
(FZK), S. Monti (ENEA), H. Aït Abderrahim (SCK-CEN) - FISA 2006 Conference on EU
Research and Training in Reactor Systems 13-16 March 2006 Luxembourg
Lead-cooled Reactors New Concepts and Applications Dr. Peter Fomichenko
Institute of Nuclear Reactors, Russian Research Center“Kurchatov Institute” Moscow
Russia - The 2004 Frédéric JOLIOT & Otto HAHN Summer School AUGUST 25 –
SEPTEMBER 3, 2004 CADARACHE, France
2007 Annual Report, GEN IV International Forum
Status Report on the Small Secure Transportable Autonomous Reactor
(SSTAR)/Lead-Cooled Fast Reactor (LFR) and Supporting Research and
Development September 29, 2006 Massachusetts Institute of Technology Ecole des
Mines de Paris Oregon State University University of California, Berkeley
The Elsy Project - L. Cinotti, G. Locatelli, H. Aït Abderrahim, S. Monti, G. Benamati,
H. Wider, D. Struwe, A. Orden – ENC 2007 – 16-20 September 2007 Bruxelles,
Belgium
Comparison Of Sodium And Lead-Cooled Fast Reactors Regarding Severe Safety
And Economical Issues Kamil Tucek*, Johan Carlsson, Hartmut Wider- Joint
Research Centre Of The European Commission Institute For Energy, Nl-1755 Zg
Petten, The Netherlands - 13th International Conference on Nuclear Engineering
Beijing, China, May 16-20, 2005
Lead-Cooled Fast Reactor Systems and the Fuels and Materials Challenges - T. R.
Allen and D. C. Crawford - Hindawi Publishing Corporation - Science and
Technology of Nuclear Installations - Volume 2007
FY2005 Ten-Year Program Plan - Appendix 4.0 - Lead-Cooled Fast Reactor – March
2005 – Idaho National Laboratory
Liquid Metal Cooled Reactors: Experience in Design and Operation, IAEA, VIENNA, 2007, IAEA-TECDOC-1569