The ALFRED Project

Abstract

The Generation IV International Forum (GIF) member countries identified the six most promising advanced reactor systems and related fuel cycle, as well as the R&D needed to establish the feasibility and performance capabilities, of the next generation nuclear energy systems known as Generation IV. Among the promising reactor technologies for fast reactors being considered by the GIF, the Lead Fast Reactor (LFR) has been identified as a technology with great potential to meet the goals of increased safety, improved economics for electricity production, reduced nuclear wastes for disposal, and increased proliferation resistance.
Ansaldo Nucleare, as coordinator of the LEADER project (Lead-cooled European Advanced DEmonstration Reactor), funded by the EC in the frame of the 7th FP, is promoting research and development on LFR. The LEADER project has the objective to design a commercially viable reactor, for large-scale electricity production (European Lead Fast Reactor - ELFR), based on lead coolant technology. The industrial deployment of ELFR requires a scaled demonstrator reactor, with the objective to demonstrate the achievement of the required safety standards, to assess economic competitiveness of lead technology, and to validate the engineering options and materials selection.
The paper presents, after a summary of the project, the main design features of the demonstrator ALFRED (Advanced Lead Fast Reactor European Demonstrator). In particular, relevance is given to the description of the Reactor configuration and the main components such as Reactor Vessel, Steam Generator, Primary Pump and Decay Heat Removal System and core configuration.
Moreover, the main challenges to the fundamental safety functions for a lead fast reactor are described and the safety provisions implemented in the design are reported. The adopted safety approach is briefly described and the main results from the preliminary safety analyses are summarized.

Full text

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The ALFRED Project
Alessandro Alemberti, Luigi Mansani, Monica Frogheri, IlieTurcu, Marin Constantin
Alessandro.Alemberti@ann.ansaldo.it,Luigi.Mansani@ann.ansaldo.it,MonicaLinda.Frogheri@ann.ansaldo.it,Ilie.Tur cu@nuclear.ro,Marin.Constantin@nuclear.ro
Ansaldo Nucleare SpA Institute of Nuclear Resaerch
Corso Perrone 25, 16159 Genova (Italy) Str. Campului, No 1 115400 Mioveni, Pitesti (Romania)
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Abstract
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The Generation IV International Forum (GIF) member countries identified the six most promising
advanced reactor systems and related fuel cycle, as well as the R&D needed to establish the
feasibility and performance capabilities, of the next generation nuclear energy systems known as
Generation IV. Among the promising reactor technologies for fast reactors being considered by the
GIF, the Lead Fast Reactor (LFR) has been identified as a technology with great potential to meet
the goals of increased safety, improved economics for electricity production, reduced nuclear
wastes for disposal, and increased proliferation resistance.
Ansaldo Nucleare, as coordinator of the LEADER project (Lead-cooled European Advanced
DEmonstration Reactor), funded by the EC in the frame of the 7th FP, is promoting research and
development on LFR. The LEADER project has the objective to design a commercially viable
reactor, for large-scale electricity production (European Lead Fast Reactor - ELFR), based on lead
coolant technology. The industrial deployment of ELFR requires a scaled demonstrator reactor,
with the objective to demonstrate the achievement of the required safety standards, to assess
economic competitiveness of lead technology, and to validate the engineering options and
materials selection.
The paper presents, after a summary of the project, the main design features of the demonstrator
ALFRED (Advanced Lead Fast Reactor European Demonstrator). In particular, relevance is given
to the description of the Reactor configuration and the main components such as Reactor Vessel,
Steam Generator, Primary Pump and Decay Heat Removal System and core configuration.
Moreover, the main challenges to the fundamental safety functions for a lead fast reactor are
described and the safety provisions implemented in the design are reported. The adopted safety
approach is briefly described and the main results from the preliminary safety analyses are
summarized.
1. INTRODUCTION
The Lead-cooled Fast Reactor (LFR) technology has a great potential to fulfill all the main goals
established by the “Generation IV International Forum” (GIF) for the next generation nuclear power
plants. The LFR is based on a closed fuel cycle for efficient conversion of fertile uranium and
management of actinides (enhanced sustainability), the inert nature of the coolant provides
important design simplification (improved economics) and allows for designing decay heat removal
systems based on well-known light water technology and passive features (increased safety).
Moreover, the reference LFR fuel (MOX) constitutes a very unattractive route for diversion or theft
of weapons-usable materials and provides increased physical protection against acts of terrorism
(Non-proliferation and Physical Protection).
At international level, with regard to lead cooled reactors, GIF considered several options: a small
transportable system of 10–100 MWe size (SSTAR – US) that features a very long core life, a
system of intermediate size (BREST 300 – Russia), and a larger system rated at about 600 MWe
(ELFR – EU), intended for central station power generation.

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In Europe, the European Commission organized the Sustainable Nuclear Energy Technology
Platform (SNETP) that, through its Strategic Research and Innovation Agenda [1] promoted the
development of fast reactors with closed fuel cycle. The Roadmap proposed by the European
Sustainable Nuclear Industrial Initiative (ESNII), includes the lead-cooled fast reactor as an
alternative technology to be developed in parallel with the sodium-cooled fast reactor.
Since the beginning, Ansaldo Nucleare has been involved in the extensive R&D programs
launched in Europe and has gathered a considerable knowledge of the lead and lead-bismuth
technologies.
In the period 1999-2001, Ansaldo Nucleare, within a group of Italian organizations, worked out a
first configuration design of a Lead-Bismuth Eutectic cooled Experimental Accelerator Driven
Systems (ADS).
The activity continued within the 5th Framework Program (FP) of the European Commission, that
funded a project named PDS-XADS (Preliminary Design Studies of an Experimental Accelerator)
where Ansaldo Nucleare was the coordinator of the studies related to the Primary System and to
the Core [2].
In the frame of the 6th FP, Ansaldo Nucleare participated to the project EUROTRANS [3] [4] for the
development and assessment of an European Facility for Industrial Transmutation (EFIT) [5] and of
the smaller eXperimental Transmutation in an ADS (XT-ADS) [6]. The resulting MYRRHA/XT-ADS
facility was further developed during the CDT project (7th FP) [7], aimed to an experimental device
that may serve both as a test-bed for transmutation and as a fast spectrum irradiation facility,
operating as a sub-critical (accelerator driven) system, and as a critical reactor [8].
The first step in the development of a Lead Cooled Critical Fast Reactor in Europe started in 2006,
when EURATOM decided to fund ELSY (European Lead cooled SYstem). The ELSY project,
coordinated by Ansaldo Nucleare, developed a very innovative pre-conceptual design of an
industrial plant for electricity production able to close the fuel cycle [9].
The LFR development continued with the LEADER project (Lead-cooled European Advanced
DEmonstration Reactor), started on April 2010 in the frame of EU 7th FP [10].
2. LEADER PROJECT OVERVIEW
The LEADER project takes into account the indications emerged from SNETP, as well as the main
goals of ESNII, and aims to the development to a conceptual level of an European Lead Fast
Reactor (ELFR) industrial size plant and of a scaled demonstrator of the LFR technology –
ALFRED (Advanced Lead Fast Reactor European Demonstrator). The project involves 17 partners
from Industry, research organizations and universities. The total effort is 502 person-months over a
period of 36 months.
The open issues and the safety concerns, emerged during the previous projects, have been
analyzed in the LEADER project and a new set of design options and safety provisions proposed,
to define the reference ELFR industrial plant configuration (600 MWe). ELFR configuration is then
used to design a low cost and fully representative scaled down demonstrator of a suitable size.
ALFRED, in the role of LFR demonstrator, will show the viability of the LFR technology for use in a
future commercial power plant, being the first link of the technology chain connected to the
electrical grid. ALFRED is designed to be as close as possible to the reference industrial size plant,
but, where needed, proven and already available solutions are adopted, although different from the
design proposed for ELFR.
3. ADVANTAGES AND CHALLENGES RELATED TO MOLTEN LEAD
The primary coolant is molten lead, which presents several favorable characteristics, such as:
 a very high boiling point (1745°C) and very low Partial Pressure. For this reason, there is no
need to pressurize the plant primary side, resulting in a easier design of the reactor vessel.

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Moreover, lead boiling scenario is practically impossible and this, together with proper
reactor vessel design, drastically reduces core voiding risk.
 It is chemically inert with air and water: this allows the elimination of any intermediate loop
and the installation of the Steam Generator Unit inside the Reactor Vessel.
 favorable neutronic characteristics (it is a low moderating medium and has a low absorption
cross-section). This gives the possibility to have a fast neutron flux even with large amount
of coolant in the core; there is no need of compact Fuel Assemblies and it is possible to
have relatively large spacing among the fuel rods, with reduced core pressure losses and
thus enhanced natural circulation capability.
 Lead has a low Polonium-210 production: pure lead is not exempt from polonium formation.
In fact, Bi-209 is produced from Pb-208, and Po-210 results from the further activation of
Bi-209; however, the rate of Po production is lower by about 4 orders of magnitude than in
the case of Lead-Bismuth. The production of Po-210 is in the order of grams. Moreover,
this very small quantity remains entrapped in the lead bulk as lead-polonide.
As for all Gen IV advanced systems, the development of the technology is associated with
research challenges. In the case of the LFR, these challenges include:
 molten lead interacts with structural materials, mainly with the mechanisms of corrosion at
high-temperature and erosion. The provisions that can be adopted to improve the
compatibility of lead and steels are:
o Operate at low temperature range (400 °C - 480°C) and maintain a controlled amount
of oxygen dissolved in the coolant.
o Select material, such as Austenitic low-carbon steels (e.g. AISI 316L), ferritic-
martensitic steels (e.g. T91), 15-15/Ti steel.
o Utilize surface coatings, e .g. by alluminisation of surfaces (with Fe-Cr-Al-Y) and
surface treatment by electron beam (GESA treatment).
o Limit coolant flow velocity to a value that cause a negligible erosion (typically 2–3
m/s).
 lead has a high melting point (327.4°C) and this can result in a risk of coolant freezing,
although this is not considered to be a safety issue but an investment protection issue. To
decrease this risk, LFR is equipped with active systems to keep the lead molten during
any normal operational conditions (e.g. during planned shutdown). In case of accidental
emergency conditions, a sufficient grace time (more than 30 min) is available for the
operator to reduce the heat removed by the Decay Heat Removal System (e.g. shutdown
one DHR loop by manually closing one valve). However, promising activities are on going
on the DHR design, in order to have the reduction of the removed heat by passive
means.
 lead is opaque and this makes the in-service inspection and handling of fuel assemblies
more difficult. For this reason, each component inside the Reactor vessel is removable
and the fuel assemblies upper end extends beyond the lead free surface in the cover gas
for refueling without the need of in-vessel machines.

4. ALFRED REFERENCE CONFIGURATION
The configuration of the ALFRED primary system is pool-type [11]. This concept permits to contain
all the primary coolant within the Reactor Vessel, thus eliminating all problems related to out-of
vessel circulation of the primary coolant (Fig. 1).
The Reactor assembly presents a simple flow path of the primary coolant, with a Riser and a
Downcomer. The heat source (the Core), located below the Riser, and the heat sink (the Steam
Generators) at the top of the Downcomer, allow an efficient natural circulation of the coolant. The

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primary coolant moves upward through the pump impeller to the vertical shaft, then enters the SG
through the lead inlet holes, flows downwards on the shell and exits the steam generator.
The free level of the hot pools inside the Steam Generator & Primary Pump units is higher than the
free level inside the Inner Vessel, the different heads depending on the pressure losses across
component parts of the primary circuit. The volume between the primary coolant free levels and the
reactor roof is filled by a cover gas plenum.
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Fig.1:ALFRED3‐DSketchandReactorBlockVerticalSections
4.1. Primary System Arrangement
The Reactor Vessel (RV) is cylindrical with a torospherical bottom head. It is anchored to the
reactor cavity from the top, by means of a vessel support. The upper part is divided in two
branches by a “Y” junction: the conical skirt that supports the whole weight and the cylindrical one,
that supports the Reactor Cover. A cone frustum, welded to the bottom head, has the function of
bottom radial restraint of Inner Vessel.
A steel layer covering the reactor pit, constitutes the Safety vessel (SV). The dimensions of gap
between the safety vessel and the reactor vessel are sufficient for the In-service Inspection tools.
The safety vessel is cooled by the same system that cools the concrete of cavity walls. This
system is inserted inside the concrete and is independent from the reactor cooling systems. This
design solution mitigates the consequences of through-wall cracks with leakage of lead: any
reactor vessel leakage is discharged into the Safety Vessel. The RV and the SV are arranged in
such a manner that in case of a reactor vessel leak, the resulting primary coolant always covers
the SG inlet and the lead flow path is indefinitely maintained.
The Inner Vessel (IV) (see Fig. 2) has two main functions: Fuel Assemblies support and separation
between hot plenum and cold plenum. It is fixed to the cover by bolts and is radially restrained at
bottom. Lead flow is guided from the FAs outlet towards the PP inlet pipes by a toroid half-ring.
Moreover, the pipes that connect the hot zone with the inlet of PP are integrated in the Inner
Vessel. The cylindrical IV has a double wall shell: the outer thick wall has a structural function,
while the inner thin wall follows the core section profile.
The Core Lower grid is a box structure with two horizontal perforated plates connected by vertical
plates. The plates holes are the housing of FAs foots and the plates distances must be sufficient to
assure the verticality of FAs. The diagrid is mechanically connected to the IV with pins (possible
removal/replacement during reactor lifetime).
The Core Upper grid is a box structure like the lower grid but stiffer. It has the function to push
down the FAs during the reactor operation. A series of preloaded disk springs press each FA on its
Fuelassembly
Reactor
Vessel
InnerVessel
PumpSteam
Generator

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lower housing. A hole is present for each disk to allow the passage of instrumentation (i.e.
thermocouples).
Upper grid
Cylinder
Lower grid
Pin
Upper grid
Cylinder
Lower grid
Pin
Fig. 2 Inner Vessel: 3-D Sketch and Axial Section
4.2. Core
The design of both ALFRED and ELFR cores has been driven by the implementation of the so
called “walk away” and “adiabatic” [12] reactor concepts.
The first concept hypothesizes the possibility to have a system able to react passively to any
initiating event, through a core/system design, which allows for enhanced negative reactivity
feedbacks and sufficient design margins to accommodate the peak excursions of critical core
parameters and to sustain the new thermal regimes for a prolonged period of time without requiring
human intervention.
The adiabatic reactor concept has been the basis for the ELFR core design and concerns the
operation of a reactor with equilibrium fuel, so that the fuel composition remains the same between
two successive loadings, ensuring the full recycling of all the actinides, with either natural or
depleted uranium as only top-up/input material and Fission Products and reprocessing and fuel
fabrication losses as output (see Fig. 3).
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Fig. 3 Scheme of a Fuel Cycle Implementing an Adiabatic Reactor
The adopted core configuration of ALFRED [13] is constituted by wrapped Hexagonal Fuel
Assemblies. It utilizes MOX as fuel and uses hollow pellets and a low active height in order to
improve the natural circulation. The total power is 300 MWth.

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The core scheme (Fig. 4) is made of 171 Fuel Assemblies (FAs), 12 CR (Control Rods) and 4 SR
(Safety Rods), surrounded by 108 Dummy Elements (ZrO2-Y2O3) shielding the Inner Vessel.
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Fig. 4: ALFRED core configuration
Each Fuel Assembly (Fig. 5) is about 8 m long and consists of 127 fuel pins, fixed to the bottom of
the wrapper and restrained sideways by grids. Tungsten deadweight (Ballast) prevents buoyancy
forces in lead. Upper elastic elements (cup springs) prevent lifting induced by hydrodynamic loads
and accommodate axial thermal expansions.
The FAs upper end extends beyond the lead free surface in the cover gas for easy inspection and
handling. In this way, it is possible to make the refueling without the need of in-vessel refueling
machines.
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Figure 5: Fuel Assembly geometry
ALFRED is equipped with two diverse, redundant and separate shutdown systems (adapted from
the one that is under investigation in the frame of the CDT-MYRRHA project):
(1) CR (Control Rod) system, used for both normal control of the reactor (start-up, reactivity
control during the fuel cycle and shutdown) and for SCRAM in case of emergency. The
Control rods are extracted downward and rise up by buoyancy in case of SCRAM. The
control mechanism pushes the assembly down with a ball screw, placed, with its motor and
resolver atop the cover (at cold temperature (<70°C)), and protected from radiation by a
shielding block. The actuator is coupled to a long rod by an electromagnet. When the
coupling electromagnet is switch off (in case of SCRAM), the absorber assembly and the rod

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are free to rise up. Control rods use a 19 pins absorber bundle, cooled by the primary coolant
flow. These pins are fitted with a gas plenum collecting the Helium and Tritium, produced by
nuclear reaction of B10
(2) SR (Safety Rod) system, is the redundant and diversified complement to the control rods for
SCRAM) only. The absorber bundle stays in the primary coolant. The rod is extracted
upward and inserted downward against the buoyancy force. The absorber gets inserted by
the actuation of a pneumatic system. In case of loss of this system, a tungsten ballast will
force the absorber down by gravity in a slow insertion.
For both systems the materials considered are B4C enriched in 10B at 90% as absorber, T91 for
the guide tube, 15-15 Ti for the clad and ZrO2 (95%) - Y2O3 (5%) for the insulator and reflector.
4.3. Steam Generator and Primary Pumps Unit
The steam generator and primary pump are integrated into a single vertical unit. Eight SG/PP units
are located in the annular space between the cylindrical inner vessel and the reactor vessel wall.
The primary pump is placed in the hot side of the steam generator, having its mechanical suction in
the hot pool inside the inner vessel. The primary coolant moves upward through the pump impeller
to the vertical shaft, then enters the SG through the lead inlet holes, flows downwards on the shell
and exits the steam generator. The pump motor is located above the reactor roof.
Each Steam Generator consists of a bundle of 542 bayonet tubes immersed in the lead vessel pool
for six meters of their length. The bayonet tube is a vertical tube with external safety tube and
internal insulating layer, composed by 4 concentric tubes (Fig. 6): slave tube, inner tube, outer tube
and outermost tube.
Fig. 6 Bayonet Tube Configuration and SG 3D Scheme
The internal insulating layer (delimited by the slave tube) has been introduced to ensure the
production of superheated dry steam: in fact, without an insulating system, the high ∆T (≈115°C)
between the rising steam and the descending feedwater would promote steam condensation in the
upper part of the steam generator.
The gap between the outermost and the outer bayonet tube is filled with pressurized helium and
high thermal conductivities particles (such as synthetic diamonds) to enhance the heat exchange
capability. In case of an external tube break, this arrangement guarantees that primary lead does
not interact with the secondary water. Moreover, a tube break can be easily detected monitoring
the Helium gap pressure.
The Primary Pump (PP) is located in the hot side of a SG. It is surrounded by the SG tube bundle
and its housing is the Steam Generator casing. The Pump is fixed to the top of SG casing by a

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bolted joint. This allows an easy removal of the component. The Primary Pump studies are in
progress. Based on analyses performed during previous LFR projects (EUROTRANS, ELSY) an
axial pump has been adopted. The PP impeller material is still an issue and test campaigns must
be performed to select the proper one. The stainless steel (SS) cannot be used because the high
speed achieved at the top of the impeller blades induces a very fast material corrosion. MAXTHAL
ceramic material has been proposed, but its reliability must be still demonstrated. An alternative
solution can be a SS impeller with a ceramic coating.
4.4. Decay Heat Removal System
The Decay Heat Removal system (DHR) consists of two passive, redundant and independent
systems, DHR1 and DHR 2, both composed of four Isolation Condenser systems (ICs) connected
to four Steam Generators (SGs) secondary side (i.e. one IC for each SG).
The system design considers the single failure criteria, since three out of four ICs are sufficient to
remove the decay heat power. The DHRs are dedicated safety systems, not used for normal
operation. The separation is achieved through placing the two DHRs in physically different
locations. A physical structural barrier or another means of protection will be placed between
adjacent IC to ensure that failure of one of them could not harm another one.
The diversity requirement has been relaxed (in any case the two DHR systems will be fabricated
by diverse manufacturers) due to the high redundancy and considering that the SG tubes bayonet
concept allows a continuous monitoring of the SG status. Both systems are completely passive,
with an active actuation through valves equipped with redundant and diverse energy sources
(batteries or locally stored energy).
Each DHR system must be ready to operate after the reactor trip in order to remove the decay heat
power, in case of unavailability of the normal path (i.e. the by-pass to the Condenser). The
actuation logic shall guarantee the actuation of DHR1 first, whereas the DHR2 shall actuate only in
case of failure of the first system. Moreover, the total number of isolation condensers called to
operate shall never overcome the four units, in order to avoid an excessive cooling of the primary
coolant leading to fluid solidification.
Each of the four independent IC sub-systems consists of (Fig. 7):
 One heat exchanger (isolation condenser), constituted by a vertical tube bundle with an upper
and lower horizontal header.
 One water pool, where the isolation condenser is immersed; the amount of water contained in
the pool is sufficient to guarantee 3 days of decay heat removal operation.
 One condensate isolation valve (to meet the single failure criteria this function shall be
performed at least by two parallel valves).
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Fig. 7 ALFRED Isolation Condenser Scheme and IC Bundle

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Each isolation condenser is connected to a Steam Generator: the upper header of the isolation
condenser is connected to the main steam line and the lower header of the Isolation Condenser is
connected to the main feed water line.
In normal operation. the isolation valve below the condenser is closed, the condenser is full of
water and no heat exchange takes place. As the IC subsystem is called to operate, the feed water
line and steam line are isolated and the condensate isolation valve opens. The sub-cooled water
stored into the IC tube bundle drains into the steam generator, due to the hydrostatic head existing
between the reactor and the isolation condenser. The IC tubes and headers become empty and
their internal cold surface starts to condensate the steam coming from the steam generator and
hence to transfer heat to the cold pool water. The steam condensation causes a pressure
reduction which calls other steam from the steam generator. The water injected into the secondary
system from the isolation condenser during the draining phase, vaporizes into the steam generator
tube bundle contributing to the secondary system pressure rise: the safety relief valves continue to
operate to reject to the atmosphere the excess of steam and to guarantee a secondary side
pressure of 195 bar(a). When the IC reaches its steady state condition and starts to remove the
thermal power from the primary coolant, the secondary side pressure rise decreases and finally
stops leading to the closure of the safety relief valves on the main steam line.
4.5. Secondary System
The Secondary system proposed for ALFRED is based on a dual turbine configuration with three
extractions in the HP turbine and three more in the LP turbine, with an axial outlet.
There shall be a reheating with steam from the first extraction and six preheaters supplied with
steam from each turbine extraction, as well as a final heater supplied with main steam. This main
steam shall be adequately throttled so that the feedwater temperature at the inlet of the steam
generator (FWTC - feedwater temperature control) can be controlled. In addition, the deaerator can
be fed from the outlet of the HP turbine.
The typical redundancy for the condensate and feedwater pumps (2x100% pumps) has also been
considered.
An auxiliary lead heating system is added. This system would work when the power cycle is not in
operation, in order to ensure the minimum temperature of the lead by transmitting heat from the
secondary system if it is needed.
A 100% turbine bypass system is included, permitting direct transfer from the reactor to the
condenser (bypass mode).
A sketch of the current ALFRED layout with forced draft cooling towers is reported in Fig. 8.
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FIG. 8 ALFRED general layout

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5. SAFETY APPROACH FOR LFR PLANT
As one of the six currently developed and analyzed Generation IV reactor systems, LFR follows
the general guidelines of the Generation IV safety concept recommendations. Among the goals for
future nuclear energy systems, improved safety and higher reliability is recognized as an essential
priority in the development and operation of nuclear power plants.
In the framework of the LEADER project, a global safety approach for the LFR reference plant has
been assessed and the safety analyses methodology has been developed [14].
The fundamental safety objectives and the Defence in Depth approach, as described by IAEA
Safety Guides, have been preserved, since they have been widely applied in the Generation II and
III nuclear power plants, characterized by a level of safety which is considered already satisfactory.
One specific challenge for Generation IV systems is to develop an approach to defence in depth
that is both consistent with the successful practices that have been used in operating reactors, and
that makes use of the improved analytical methods that have come to be available to focus on
defence in depth design provisions in such a way as to cost-effectively optimize the value of that
defence in depth. For Generation IV systems, the goal will be to apply defence in depth in a
manner that explicitly takes into consideration uncertainties based on their systematic assessment.
The ideal outcome will be a design that optimizes both capital costs and safety by applying
defence in depth where it will have the desired effect, but not to “over-design” in a way that adds
cost but provides little additional value in safety.
The recommendation of the Risk and Safety Working Group (RSWG1) has been taken into
account, in particular:
 safety is to be “built-in” to the fundamental design rather than “added on”;
 full implementation of the Defence-in-Depth (DiD) principles:
 Exhaustive: the identification of initiating events used to design the safety architecture
should be as exhaustive as possible;
 Progressive: without that, “short” sequences can happen for which, downstream from the
initiator, the failure of a particular provision entails a major increase, in terms of
consequences, without any possibility of restoring safe conditions at an intermediate
stage;
 Tolerant: no small deviation of the physical parameters outside the expected ranges, can
lead to severe consequences (i.e. rejection of “cliff edge effects”);
 Forgiving: availability of a sufficient grace period and the possibility of repair during
accidental situations;
 Well-balanced: no sequence participates in an excessive and unbalanced manner to the
global frequency of the damaged plant states;
 “risk-informed” approach - deterministic approach complemented with a probabilistic one;
 adoption of an integrated methodology that can be used to evaluate and document the safety
of Gen IV nuclear systems - Integrated Safety Assessment Methodology (ISAM). In particular
the Objective Provision Tree (OPT) tool is the fundamental methodology used throughout the
design process. The OPT is a top-down method with a tree structure which, for each level of
DiD and for each safety objective/function, identifies the possible challenges to the safety

1 The RSWG was formed in the frame of Generation IV International Forum (GIF) to promote a homogeneous and
effective approach to assure the safety of Generation IV nuclear energy systems

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functions, the plausible mechanisms which can materialize these challenges and the provided
provision(s) to prevent, control or mitigate the consequences of the challenges/mechanisms
Finally, although The European Utility Requirements (EUR) have been developed for LWR plants,
the General requirements regarding the safety approach and the quantitative Safety objectives
have been adopted, in order to have a “minimum” target (EUR quantitative objectives are more
stringent than those of the current plants) to be pursued in the development of a Generation IV
plant.
6. ALFRED SAFETY ANALYSIS: SUMMARY OF THE RESULTS
6.1. DBC Transients
The results of the performed preliminary safety analysis [15] show that ALFRED meets the DBC
acceptance criteria.
Calculated peak pin temperatures do not exceed 700 °C, thus no clad failures (creep rupture) are
expected, even during the loss of AC power (PLOOP) and loss of all primary pumps (PLOF)
transients. This is mainly due to the large thermal inertia of the Pb-cooled primary system and the
relatively high Pb natural convection core mass flow rate, observed at the initial stage of the
transient. In the long-term, in case of PLOOP or PLOF transients, following reactor trip, core decay
heat can be passively and safely removed indefinitely by the DHR-1 system without the need of
any AC power.
Large margin to clad failure (rupture) of the peak pins was confirmed also during the simulated
peak power following a 70% FA flow blockage transient. In this case, reactor safety is ensured by
the reactor protection system, shutting down the reactor, as well as by the DHRS, efficiently
removing the decay heat from the reactor primary cooling circuit. A reactor trip delay, even of 10
sec, does not lead to clad failure.
During the protected FW temperature decrease transient to 300 oC, apart from an issue of Pb-
freezing at the outlet of the SG several hours into the transient, no safety related issues are
observed. During this time period, the operator has sufficient time to deactivate sub-systems of the
DHR-1 system, in order to prevent Pb-coolant freezing.
As a general conclusion, no relevant, immediate safety issues have been identified during the
performance of all the simulated protected DBC transients, aside of potential Pb-freezing several
hours into the transient. Reactor safety is assured by the reactor protection system, shutting down
the reactor, in conjunction with a heat removal system (DHRS), removing tightly controlled
amounts of decay heat from the reactor primary cooling circuit to assure prevention of freezing of
the Pb-coolant at any location of the primary cooling circuit.
The analysis indicated that ALFRED is a very forgiving plant design, and there is an extended time
margin (grace time) of several hours for possible manual operator intervention even under worst
accidental conditions (potential of Pb-freezing in the long term, in case of uncontrolled decay heat
removal by the DHRs). However, design modifications to the DHRs are under study in order to
avoid the need of operator intervention to prevent freezing, thus warranting an infinite grace time
(DHRs with indefinitely passive operation).
6.2. DEC Transient
The analysis of the representative DEC transients [16] for ALFRED has highlighted very good
intrinsic safety features of the reactor design.
In particular, the results of the unprotected transients underline that the reactor can be maintained
in a safe and controlled state, even under the unlikely accident conditions, that include the failure
of the reactor scram, thanks to:
 The establishment of enhanced and stable natural convection in the primary cooling circuit
following the loss of the primary pumps,
 The dominant negative reactivity feedback effects obtained by optimizing the neutronic core
design,

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 The large thermal inertia of the primary cooling system,
 The passive and efficient operation of the DHR system for decay heat removal.
In all simulated unprotected transients the core temperatures are maintained well below the safety
limits to be considered for DEC accidental transients. In particular:
 In all simulated transients there is a very large margin to coolant boiling, since the coolant
temperature is always at least 900 °C below the lead boiling point (1740 °C);
 No clad failure is predicted in any of the simulated transients, unless in case of an
undetected FA blockage greater than ~85%, which might be excluded by design (there are
many orifices for coolant ingress at the FA inlet), and in case of the very unlikely
ULOF+ULOHS event, when the time-to-failure reduces down to a few hours, but still leaving
enough grace time for corrective operator actions;
 Fuel melting is excluded in all simulated transients except for local fuel melting (in the fuel
pellet center) in the hottest FAs in case of reactivity insertion involving core compaction or
core voiding, due to passage of steam bubbles transported at the core inlet following the
steam generator tube rupture event;
As a general conclusion, no relevant safety issues have been identified for ALFRED in case of
representative DEC events. In particular, the ULOF transient can be accommodated without the
need of corrective operator actions. Finally, the analysis of DEC transients for the lead-cooled
ALFRED design has demonstrated the extremely forgiving nature of this plant when compared to
other similar plant designs.
7. LFR RESPONSE TO EXTREME NATURAL EVENTS
In the light of the events which occurred at Fukushima, extreme natural events challenging the
plant safety functions and leading to a severe accident must be considered, in order to assess the
plant safety margins during extreme situations. In particular, combination of initiating events and
failures of design provisions resulting in the sequential loss of the lines of defense has to be
assumed, in a deterministic approach, irrespective of the probability of this loss.
The safety analyses performed for ALFRED, together with the design features developed on the
basis of the identified challenge mechanisms and the inherent favorable characteristics of Lead
coolant, allow the following considerations:
 The response of the plant to earthquakes has been enhanced, as concerns the structural
integrity, with the adoption of 2D seismic isolators. The potential reactivity insertion due to a
core compaction has been anticipated, by the adoption of wrapped hexagonal Fuel
assemblies, with lateral restrains.
 In case of loss of secondary system, the Decay Heat Removal is assured by two
independent and redundant systems, connected to the secondary side (SG steam line).
 The DHR systems is passive. Only its actuation (through valves alignment) is active, but
there is a backup energy source. Hence, the Station Blackout does not represent a concern
for the DHR system.
 The safety analyses performed , have addressed the case with Loss of Primary Flow +
Heat Sink (i.e. Station Blackout) without DHR systems available, demonstrating that, even
in these conditions, the fuel and cladding temperatures do not represent a concern.
 The lead coolant is characterized by high thermal inertia (as reflected by the large coolant
mass times heat capacity in the core region and in the primary system pool) and by a very
high lead boiling point. Moreover, lead density is slightly higher than that of the oxide fuel,
resulting in an enhancement of the fuel dispersion with respect to fuel compaction. These
favorable intrinsic lead characteristics make the complete core melt practically impossible.
 In the very unlikely event of a Fukushima like scenario leading to the loss of all heat sinks
(both DHR and secondary systems), the heat can be extracted injecting water in the reactor

13
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cavity between the reactor and safety vessels, while in case of reactor vessel breach the
decay heat can still be removed by the same system that cools the concrete of cavity walls.
Such very ultimate provisions in a Fukushima like scenario are possible because Lead is
chemically inert with air and water so that fire fighters or dedicated rescue troops can flood
the reactor using water.

8. THE ALFRED ROADMAP
The main drawback facing the industrial deployment of a LFR fleet in Europe is the lack of
operational experience gathered so far. The aim at filling the technology gap in Europe requires the
setting up of a complete R&D roadmap [17].
Besides all the existing and planned EU lead labs, as well as the further facilities envisaged for
fulfilling the technology gaps in thermal/hydraulics, materials development and corrosion testing, a
complete set of nuclear facilities is needed to bring the LFR technology to industrial maturity.
In this frame, the following facilities are well fitting: GUINEVERE, the zero-power facility operating
in Mol (Belgium) since 2010; MYRRHA, the irradiation facility to be built and operated in Mol and
the European training facility ELECTRA, planned for realization in Sweden. In this sense, the
approach to the industrial maturity is based on the concept of progressive up scaling, assuming the
zero-power GUINEVERE as the starting point.
The LFR development requires at first a small Demonstrator reactor (ALFRED) for proving the
viability of reliable electricity production for LFR systems. A Prototype reactor (PROLFR) is then
envisaged, for testing the scaling laws at an intermediate step – according to a common approach
focused both on plant size and representativeness of the target reference system – before moving
to the First-Of-A-Kind (FOAK) representative of a commercial ELFR fleet (Fig. 9).

Fig. 9 Overall Roadmap for Deployment of a European LFR
The realization of ALFRED will include several phases (2010-2025), as shown in Fig. 10. The first
step is identified in the set-up of an international consortium (2013), while the choice of a dedicated
site for construction has been already performed (Pitesti in Romania).
In parallel, several design activities should be running: basic design, siting and pre-licensing in the
period 2013-2016 and detailed design and licensing in the period 2016-2019. Finally construction
of components and civil engineering, on site assembly and commissioning (2019-2025).

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The completion of the design phase is presently planned for 2019 to be able to exploit the
synergies with the design of MYRRHA. In parallel, the support R&D program will provide, during
the 2011-2018 period, the necessary answers to the remaining technical challenges.
Fig. 10 ALFRED Implementation Schedule
9. CONCLUSIONS
The paper presented a summary of the LEADER (Lead-cooled European Advanced
DEmonstration Reactor) project, funded by the European Commission in the frame of the 7th FP,
aimed to the development to a conceptual level of a European Lead Fast Reactor(ELFR) industrial
size plant and of a scaled demonstrator of the LFR technology ALFRED (Advanced Lead Fast
Reactor European Demonstrator). In particular, the main design features of the demonstrator
ALFRED are described together with the main safety characteristics. The LFR roadmap, aimed to
arrive at the construction of the FOAK ELFR with the intermediate step foreseen (e.g. construction
of a demonstrator ALFRED and a prototype PROLFR), are also showed.
Acknowledgements
The authors acknowledge the European Commission for funding the LEADER project in its 7th
Framework Programme. Acknowledgment is also due to all the colleagues of the participant
organizations for their contributions in many different topics.

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