LEAD-COOLED FAST REACTOR

Lehrstuhl für Nukleartechnik - Technische Universität München

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www.ntech.mw.tum.de

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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

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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.

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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.

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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

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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 $;

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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)

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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

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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

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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

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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.

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Figure 1.4: General view of RBEC reactor

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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.

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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:

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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

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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

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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

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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.

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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.

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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.

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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.

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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.

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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.

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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