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Home : Generation IV Technology : Systems : Lead-Cooled Fast Reactor

The Lead-Cooled Fast Reactor (LFR) system features a fast-spectrum lead or lead/bismuth eutectic liquid-metal-cooled reactor and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides.

The LFR system has excellent materials management capabilities since it operates in the fast-neutron spectrum and uses a closed fuel cycle for efficient conversion of fertile uranium. It can also be used as a burner to consume actinides from spent LWR fuel and as a burner/breeder with thorium matrices. An important feature of the LFR is the enhanced safety that results from the choice of molten lead as a relatively inert coolant. In terms of sustainability, lead is abundant and hence available, even in case of deployment of a large number of reactors. More importantly, as with other fast systems, fuel sustainability is greatly enhanced by the conversion capabilities of the LFR fuel cycle.

The LFR was primarily envisioned for missions in electricity and hydrogen production, and actinide management. Given its R&D needs in the areas of fuels, materials, and corrosion control, a two step process leading to industrial deployment of the LFR system has been envisioned: by 2025 for reactors operating with relatively low primary coolant temperature and low power density; and by 2035 for more advanced designs. The preliminary evaluation of the LFR concepts considered by the LFR Provisional System Steering Committee (PSSC) covers their performance in the areas of sustainability, economics, safety and reliability and proliferation resistance and physical protection.

The LFR concepts that are currently being designed are two pool-type reactors (Table1):

  • the Small Secure Transportable Autonomous Reactor (SSTAR), developed in the USA and
  • the European Lead-cooled System (ELSY), developed by the EC.

Table 1. Key Design data of GIF LFR concepts

Parameters
SSTAR
ELSY
Power [MWe]
19.8
600
Conversion Ratio
~1
~1
Thermal efficiency [%]
44
42
Primary coolant
Lead
Lead
Primary coolant circulation
(at power)
Natural
Forced
Primary coolant circulation
for direct heat removal (DHR)
Natural
Natural
Core inlet temperature [°C]
420
400
Core outlet temperature [°C]
567
480
Fuel
Nitrides
MOX, (Nitrides)
Fuel cladding material

Si-Enhanced Ferretic/Martensitic
Stainless Steel

T91 (aluminized)
Peak cladding temperature [°C]
650
550
Fuel pin diameter [mm]
25
10.5
Active core dimensions
Heigh/equivalent diameter [m]
0.976/1.22
0.9/4.32
Primary pumps
-
N°8, mechanical,
integrated in the SG
Working fluid
Supercritical CO²
at 20 MPa, 552°C
Water-superheated steam at 18 MPa, 450°C
Primary/secondary heat transfer system
Four Pb-to-CO² HXs
Eight Pb-to-H2O SGs
Direct heat removal (DHR)
Reactor Vessel Air Cooling System
+
Multiple Direct Reactor Cooling Systems
Reactor Vessel Air
Cooling System
+
Four Direct Reactor Cooling Systems
+
Four Secondary Loops Cooling Systems

 

It should be noted that the objective of designing LFR with the high mean core outlet coolant temperatures required for the generation of hydrogen by thermo-chemical processes, could not been addressed simultaneously with the two-track design approach of the systems indicated above, owing to the required longer term R&D necessary for the development of new high-temperature materials that will be needed to provide corrosion resistance with lead as the coolant; this objective will be addressed at a later stage, depending on the success of the nearer term technology demonstration stage, that has been given priority.

The SSTAR is a small factory-built turnkey plant operating on a closed fuel cycle with very long refuelling interval (15 to 20 years or more) cassette core or replaceable reactor module. The current reference design for the SSTAR in the United States is a 20 MWe natural circulation reactor concept with a small shippable reactor vessel (Figure 1). Specific features of the lead coolant, the nitride fuel containing transuranic elements, the fast spectrum core, and the small size combine to promote a unique approach to achieve proliferation resistance, while also enabling fissile self-sufficiency, autonomous load following, simplicity of operation, reliability, transportability, as well as a high degree of passive safety. Conversion of the core thermal power into electricity at a high plant efficiency of 44 % is accomplished utilizing a supercritical carbon dioxide Brayton cycle power converter.

 

Figure 1. - Conceptual 20 MWe (45 MWt) SSTAR system

The initial design of ELSY is almost complete. The next step in its development is the R&D testing of several design innovations, in order to start with confidence, the detailed engineering design of a reduced-scale demonstration facility.

The ELSY reactor (Figure 2) is rated at 600 MWe. This mid-size rating is the result of the fact that plants of the order of several hundreds MWe are most economically attractive for addition to the European interconnected grids. In addition, a larger plant would require an increase mass of the lead coolant and would entail increased mechanical loads on the reactor vessel and its supporting structure.

The choice of a mid-size reactor power suggested the use of forced circulation to shorten the reactor vessel thereby avoiding excessive coolant mass and alleviating mechanical loads on the reactor vessel.

Thanks to the favorable neutron characteristics of lead, the fuel rods have been spaced further apart than in the case of previous fast-neutron cores. This and the innovative steam generators with flat spirals tube bundle enable the design of a low pressure loss primary loop. The needed pump head, in spite of the higher density of lead, could, therefore, be kept low (on the order of two bars) with reduced requirement of pumping power.

Because of the predicted low primary system pressure loss and the favorable heat transfer properties of lead, decay heat can be removed by natural circulation in the case of loss of station service power (LOSSP).

The proposed thermal cycle of 400 °C at core inlet and only 480 °C at core outlet enable key advantages in terms of use of currently-available structural steels, reduced corrosion and reduced creep thereof, and milder thermal transients.

In terms of efficiency of electricity energy generation, the designers have achieved almost the same thermal efficiency as the Na-cooled SPX1, in spite of the 62 K lower mean core outlet temperature.

The potential of specifying higher operating temperatures of the primary cycle, owing to the low vapour pressure and very high boiling point of lead, depends on the qualification of suitable structural materials, the development of which may prove a long-term task, and has not been included in the near-term development of the LFR. However, the potential for future high temperature operations remains an attractive feature of the LFR.

Priority has been given to the designer’s goal of demonstration of the technical feasibility of the LFR within a relatively short time frame, with features such as a MOX-fuel core self-sustaining because of a conversion ration of about 1 and being adiabatic to (i.e. burner of) the self-generated MA. Development of the LFR to the more ambitious goals of high temperature operation and burning capability of MA beyond the self-generated MA will be considered, but will be pursued in detail in a future stage, depending on R&D and design achievements, and budget.

Figure 2 - ELSY reference configuration. Status at the end of 2008.

Advantages and challenges

The main advantages of the LFR system are its expected fuel efficiency, its capabilities in terms of nuclear materials management (thereby mitigating proliferation risks) and the reduced production of high-level radioactive waste and actinides.

The main features that the members have identified in order to achieve the Generation IV goals are summarized in Table 2. These features are based either on the inherent features of lead as a coolant or on the specific engineered designs.

Table 2  LFR potential performance against the four Goal Areas and the eight Goals for Generation IV.

 

Generation IV Goal Areas
Goals for Generation IV Nuclear Energy Systems

Goals achievable via

Inherent features of Lead
Specific engineered solutions

Sustainability

Resource utilization
  • Lead is a low moderating medium
  • Lead has low absorption cross-section
  • This enables a core with fast neutron spectrum even with a large coolant fraction
  • Conversion ratio close to 1
Waste minimization and management
  • Great flexibility in fuel loading including homogeneously diluted MA

Economics

Life cycle cost
  • Lead does not react with water
  • Lead does not burn in air
  • Lead has a very low vapor pressure
  • Lead is inexpensive
  • Reactor pool configuration
  • No intermediate coolant loops
  • Compact primary system
  • Simple design of the reactor internals
  • Supercritical water (high efficiency)
Risk to capital
(Investment protection)
 
  • Small reactor size
  • Potential for in-vessel replaceable components
  • Long refuelling cycle

Safety
and
Reliability

Operation will excel in safety and reliability

Lead as:

  • Very high boiling point
  • Low vapor pressure
  • High shielding capability for gamma radiation
  • Good fuel compatibility and fission product retention
  • Primary system at atmospheric pressure
  • Low coolant ΔT between core inlet and outlet
Low likelihood and degree of core damage

Lead as:

  • Good heat transfer characteristics
  • High specific heat and thermal expansion coefficient
  • Core with inherent negative reactivity feedback
  • Large fuel pin pitch
  • Natural circulation cooling (small system)
  • Decay Heat Removal (DHR) in natural circulation
  • Primary pumps in the hot collector (moderate - or large - size system)
  • DHR coolers in the cold collector
No need for offsite emergency response
  • Lead density is close to that of fuel (considerably reduced risk of re-criticality in case of core melt)
  • Lead retains released fission products
 

Profiferation Resistance
and
Physical Protection

Unattractive route for diversion of weapon-usable material
  • Lead system neutronics enables long core life
  • Small system features sealed, long-life core
  • Use of a MOX fuel containing MA increases proliferation resistance
Increased physical protection against acts of terrorism
  • Primary coolant chemically compatible with air and water operating at ambient pressure
  • Simplicity in design
  • Independent, redundant and diversified DHR loops
  • No use of reactive or flammable coolant materials

 

Overview of key challenges for the LFR is provided in table 3.

Table 3. Key challenges of the LFR design.

General issue

Specific issue

Proposed strategy

Corrosion in Lead

Tendency for material corrosion with increasing temperature

Mean core outlet temperature for the large plant is limited to 480°C¹
Dissolved oxygen provides barrier against corrosion

Reactor vessel

Temperature limited by design to 400°C

Fuel cladding

Use of aluminized surface treatment of steels

Reactor internals

Dissolved oxygen control

SG tubes

Use of aluminized steels to avoid lead pollution and heat transfer degradation

Pump impeller degradation²

Use of innovative materials

Seismic design

Challenge related to the large mass of lead

Use of 2D seismic isolators + short vessel design

SGs are installed inside the reactor vessel
with risk of water ingress in lead in case of SGTR accident

Rupture of the SG collectors in lead

Eliminated by design

Steam entrainment in the core in case of SGTR

Excluded by design

Pressure waves inside the primary system in case of SGTR

Harmless by specific design features

DHR

Diversification, reliability and passive operation required³

Diversification and reliability by means of use of both atmospheric air and stored water

Refuelling in lead

High temperature makes refuelling difficult⁴

Access to fuel assemblies is in cold cover gas

¹ The small system operates at a higher temperature but because of the use of natural circulation cooling the erosive effect of lead is reduced
² Pump impeller problem is not characteristic to small system because of the use of natural circulation cooling
³ In case of small system a simple and reliable RV air cooling system is sufficient to remove decay heat
⁴Small system feature sealed core without refueling or complete replacement as a cassette

Most challenges have been positively addressed by the conceptual ELSY design configuration as of the end of 2008, but the challenge remains of the follow-on design of a very high temperature reactor, operating beyond 550°C, the design of which has not yet been addressed, mainly because of outstanding information about corrosion resistant, high-temperature materials.

GIF progress in 2008

The LFR R&D development plan incorporates two tracks of development leading to a single joint demonstration facility by 2020. Separate designs for a small, transportable LFR with a long core life and a moderate-sized power plant will be researched in the demonstration facility. The LFR system research plan, which sets out the research required in the system design, fuel and lead technology and materials, was updated in the course of 2008.

Recent LFR research papers and links

Cinotti L., et al., “The ELSY Project”, Paper 377, Proceeding of the International Conference on the Physics of Reactors (PHYSOR), Interlaken, Switzerland, 14-19 September, 2008.

L. Cinotti et al, The Potential of the LFR and the ELSY Project, 2007 International Congress on Advances in Nuclear Power Plants (ICAPP '07).

Y. H. Yu, H. M. Son, I. S. Lee, K. Y. Suh, Optimized Battery-Type Reactor Primary System Design Utilizing Lead, Paper 6148, 2006 International Congress on Advances in Nuclear Power Plants (ICAPP'06).

I.S. Hwang, A Sustainable Regional Waste Transmutation System: P E A C E R, Plenary Invited Paper, 2006 International Congress on Advances in Nuclear Power Plants (ICAPP'06).

W. J. Kim, T. W. Kim, M. S. Sohn, K. Y. Suh, Supercritical Carbon Dioxide Brayton Power Conversion Cycle Design for Optimized Battery-Type Integral Reactor System, Paper 6142, 2006 International Congress on Advances in Nuclear Power Plants (ICAPP'06).

A. V. Zrodnikov, G. I. Toshinsky, O. G. Komlev, Yu. G. Dragunov, V. S. Stepanov, N. N. Klimov, I. I. Kpytov, and V. N. Krushelnitsky, Use of Multi-Purpose Modular Fast Reactors SvBR-75/100 in Market Conditions, Paper 6023, 2006 International Congress on Advances in Nuclear Power Plants (ICAPP'06).

L. Cinotti, C. Fazio, J. Knebel, S. Monti, H. Ait Abderrahim, C. Smith, K. Suh, LFR (2006)

“LFR ‘Lead-Cooled Fast Reactor’", Proceedings of FISA 2006, EU Research and Training in Reactor Systems, Luxembourg, 13-16 March 2006

E-mail contact: [email protected]

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