Lead-Cooled Fast-Neutron Reactor

(BREST) (APPROACHES TO THE CLOSED NFC)

Yu.G.Dragunov,

V.V. Lemekhov,

A.V. Moiseyev,

V.S. Smirnov

Joint Stock Company (JSC) “N.A.Dollezhal Research

and Development Institute of Power Engineering”

INPRO Dialog-Forum,

IAEA HQ, Vienna, Austria, May 26-29 2015

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Lead-Cooled Fast-Neutron Reactor (BREST) (APPROACHES TO THE CLOSED NFC)

OUTLINE:

1. Preamble: Lead-cooled fast reactors

2. BREST–OD-300: Main goals of development, state-of-art

3. BREST–OD-300: Natural Safety principles

4. Closed Nuclear Fuel Cycle

5. Back-End of the NFC, Radiation Equivalence Principles

6. Conclusion: prospects, problems, collaboration

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FAST NEUTRON REACTOR WITH HEAVY METAL COOLANT

An comprehensive analysis of the innovative reactor technologies

of a new generation under consideration in Russia and

elsewhere shows that the concept of a fast-neutron reactor with

a heavy liquid-metal coolant meets higher safety and fuel supply

requirements.

Namely these features formed the basis of respective pioneer

reactor designs in Russia and later adopted in Europe (ELCY,

ALFRED-FALCON, MYRRHA), as well as in the USA, Japan,

China and South Korea.

The recognition of the fact that heavy metals are prospective as

coolant materials has been reflected in quite an active

international collaboration, primarily as part of IAEA programs

(INPRPO and others), GENERATION IV, etc.

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MAIN GOALS OF TEHNOLOGY

The exclusion of severe accidents of nuclear power plants (reactivity,

loss of cooling, fires, explosions), requiring the evacuation of the

population;

The closed NFC circuit to fully exploit the energy potential of uranium

feedstock;

Back-end of NFC: a consistent approach to radiation equivalence of

definitively buried RAW with respect to the originally used natural

uranium raw materials;

Technological strengthening of non-proliferation :

lack of separation U and Pu when reprocessing spent fuel,

the rejection of U blanket with Pu breeding, and

non-involvement of enriched uranium in reactor loading (i.e.

no uranium enrichment later);

Ensuring the competitiveness of nuclear power in compared with other

types of power generation

•/Dragunov Yu.G., et al, Atomnaya Energiya, v. 113, -1, 2012, pp.58-65/

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THE REACTOR BREST-OD-300:

GOALS AND OBJECTIVES OF CREATION

Goal – practical confirmation of realization of the “Natural Safety”

concept of the lead-cooled fast reactor, operating in NPP mode with

closed NFC.

Objectives:

Life experience in all stages of the life cycle for commercial power units,

built according to the chosen concept

Complete fuel breeding (equilibrium mode) for self-sustaining

Confirmation of exclusion of accidents caused by reactivity and

accidents with loss of coolant, requiring evacuation of the population, in

the imposition implemented multiple bounce for internal reasons

Gaining experience in NPP operating in the closed nuclear fuel cycle

•/Dragunov Yu.G., et al, Atomnaya Energiya, v. 113, -1, 2012, pp.58-65/

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NPP: DESIGN CONCEPT

For lack of reactivity margin enough for realization of severe

reactivity accident.

Integral-type arrangement of the first contour to avoid output of

coolant outside the reactor vessel, to eliminate lost of coolant.

Using of low-activated coolant with high enough boiling

temperature, without rough interaction with water and air in the

case of depressurizing of the contour.

Realization of full breeding of fuel within the active zone solely,

burning of the long-lived actinides.

Simplifying of the safety systems due to physical features of used

materials and design methods.

MCP

Steam generator Vessel Core

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Thermal power, MW 700

Electric power, MW 300

Steam production rate, no less than, t/hour 1480

Coolant of the first contour Lead

Gas pressure above the lead level:

- exceed, MPa

- maximal, MPa

0,003-0,008

0,02

Average temperature of the lead coolant on

the active zone entry/ exit, °С 420/540

Average temperature of the lead coolant on

the steam generator entry/ exit, °С 340/505

Loop number 4

FA number in the active zone 169

Active zone height, mm 1100

Fuel load, t 20,6

Fuel campaign, years 5

Burn-up of unloaded fuel

(maximum/ average), % HM. 9,0/5,5

Collector of SACR

The BREST: key components and technical characteristics

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Use of multilayer reinforced concrete with bimetallic facing and

coolant with high freezing point – negligible lost of lead through diffusion

of lead into concrete in the case of depressurizing of facing.

Absence of lock fittings in the first contour – impossible to break

circulation.

The coolant circulation scheme with the over-fall of free levels – a

guaranteed prolongation of circulation under lost of power supply.

The emergency coolant system with natural circulation, transferring

heat directly from the first contour to the final absorber – atmospheric air.

EXCEPTION TO LOSS OF COOLING

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The most severe failures of normal operation

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The most severe failures of normal operation

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The most severe failures of normal operation:

unplanned full lost of electric power supply,

plus failure of two stop systems at once

Тclad

Тout AZ

Тin AZ

Тin SG Тout SG

Т, С

Time, s

Тfuel

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The most severe failures of normal operation

13

The most severe failures of normal operation

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The most severe failures of normal operation:

unplanned output of the absorber rods

Тclad

Тout AZ.

Тin AZ.

Тin SG Тin SG

Т, С

Time, s

(MCP stop due to temperature,

than PFBS,

than passive safety system-Т)

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BREST-OD-300: ACTIVE ZONE

1. Central zone FAs;

2. Peripheral zone FAs;

3. Active CPS rods;

4. Active-passive CPS rods;

5. Shim rods;

6. Automatic control rods;

7. PFBS block;

8. Removable reflector block

Fuel operation in the

BREST reactor

Fuel cooling

(1 year)

Fuel refabrication

Makeup by natural or

depleted uranium

Fuel regeneration

Waste

FUEL CYCLE FLOW CHART

(V.S. Smirnov, International Workshop on Innovative Nuclear Reactors Cooled by Heavy liquid metals: Status

and Perspectives, Pisa, April, 2012) 16

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Use of mixed uranium-plutonium nitride fuel with high density and

high thermal conductivity and low moderating coolant ensure

breeding of the fissionable materials in the active zone.

Full reproduction of fission materials in the active zone (BRA ~1,05)

allows not to have a reactivity margin to burnout and, accordingly,

the margin for overclocking (including cold state), leading to severe

accidents requiring evacuation of the population.

BREEDING RATIO BR ~1,05

CLOSED FUEL CYCLE

The ultimate objectives of the BREST-OD-300 project include demonstration

of not only the expected physical and operational characteristics and intrinsic

safety of this installation as, but also its capability of operating in a closed

cycle mode with an equilibrium fuel system.

Equilibrium mode of fuel supply means that the reactor operates with

complete reproduction of the fissile nuclides in the reactor core (breeding

ratio ≈1) and fuel is recycling through the extra-reactor facilities –

components of the closed fuel cycle complex.

By this mode, the weights and isotopic compositions of Pu and MА in charged

(fresh) and discharged (spent) fuel would be virtually the same, and

ultimately, the only one burnt component would be 238U, whose mass would

be replenished every time as new fuel is produced.

V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear

Technologies and Environment, № 1'2012 18

BASIC DIAGRAM OF BREST FUEL REGENERATION

(V.S. Smirnov, International Workshop on Innovative Nuclear Reactors Cooled by Heavy liquid metals: Status

and Perspectives, Pisa, April, 2012)

Expected initial load:

• mixed mono-nitride: ~13.2% Pu in U-Pu

• Plutonium isotopic composition – corresponds to Pu, extracted from SNF of VVER

after 25 years of cooling.

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CLOSED FUEL CYCLE

Environmentally safe closing of the fuel cycle would be achieved through

utilization of specific fuel recycling and refabrication technologies that only

require relatively coarse treatment of spent fuel to remove fission products,

adding depleted uranium to the treated fuel mix (U-Pu-minor actinides),

nitration and fabrication of new fuel.

Irradiation time up to planed average burn-up (cca 8% HA) is 5 year.

After discharge from the core, assemblies with spent fuel would be placed in

at-reactor storage, cooling for 1 year and then being shipped to reprocessing

plant.

Spent fuel reprocessing and new assemblies fabrication take the next 1 year.

V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear

Technologies and Environment, № 1'2012 20

Equilibrium fuel mode presumes stability of reactivity during fuel burning

between refueling (during the cycle), within the effective share of delayed

neutrons (βeff)

V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear

Technologies and Environment, № 1'2012

EQUILIBRIUM MODE OF FUEL SUPPLY

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1.0

0 300 600 900 1200 1500 1800 2100 2400 2700 3000 3300 3600 3900 4200 4500 4800 5100 5400

Кампания реактора, эф. сут

ОЗР, βэф

Reactivity margin,

βeff

Time, days

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POWER PRODUCTION COMPLEX IN PLAN

V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear Technologies

and Environment, № 1'2012 22

NPP with reactor BREST-OD-300, reprocessing plant and long term storage

on site.

RADIATION EQUIVALENCE PRINCIPLE

"The principle of “Radiation Equivalence" –

Balance radioactivity (taking into account the dangers of both the

biological impacts and the natural migration) between the hazard of

natural uranium used to produce an energy in a closed power system

and the hazard of long-lived high-level radioactive waste, elaborated by

the reactors.

Balance of radioactivity in a closed power system with BREST-

type reactors can be achieved:

on the base of in-reactor transmutation of actinides and the

extraction and

controlled long-term (other of 300 years) storage of other high-

level waste (cooling - to reduce the activity of thousands times) before

their final disposal.

(A.V.Lopatkin, V.V.Orlov, et al. Fuel cycle of BREST reactors. Solving of RAW and non-proliferation problems,

ICONE 11-36405) 23

Potential biological hazard (ingestion) of irradiated fuel

(A.V.Lopatkin, V.V.Orlov, et al., N-Novgorod, 2007)

Time after recovery, years

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Radio-equivalence is achieved

after 300 years storage provided

extraction of Am with residual in

the waste is less than 1%

25 25 (E.V. Spirin, R.M. Alexakhin, S.I. Spiridonov, The radiation balance of the spent nuclear fuel according to the criteria of

impact on human health and the environment. XLIII radio-ecological reading in memorial of V.M. Klechkovsky)

Assessment of optimal parameters for reprocessing of BREST-OD-300 INF

by criteria for the impact on the population and the biota on all potential paths

The total dose of external and internal exposure of Man by actinides

of BREST-OD-300 INF provided the residual content of U, Pu, Np - 0,1%:

- 1 ton of natural U (1);

hereinafter:

- the residual of Am in LLHL RAW - 100% (2); 10% (3); 3% (4); 1% (5); 0,3% (6) and 0,1% (7).

Dose, Zv

Time, year

TECHNOLOGICAL SUPPORT OF NON-PROLIFERATION

•All FAs of the core contain the same mass of Pu.

• No uranium blankets and no breeding of weapon-grade Pu, because blankets

do not needed.

•No need to recover Pu for fabrication of reloading fuel (it is suffice to separate

fission products and add depleted U). Hence, reprocessing may be used, but it

is not suitable for Pu recovery.

•No need for U enrichment.

•Surplus Pu is used as part of U-Pu mixture for fabrication of the first loads for

new reactors

•Reprocessed fuel is partially cleaned from fission products (recycling fuel

contains about 10-2FPs present in spent fuel) and incorporates minor actinides,

which makes fuel highly radioactive (as radiation barrier to fuel thefts). As

projected, reactors burn 238U added into fuel at refabrication.

Pu is part of fuel and recycles in the closed cycle as part of highly active fuel

(combustion catalyst for 238U ) (A.V.Lopatkin et al., 26

COMPETITIVE ABILITY

A plant with a BREST-type reactor is expected to be economically competitive

owing to the simpler design of the facility and its safety systems, as well as to

efficient utilisation of nuclear fuel and generated heat.

Low lead pressure in the circuit allows using an integral configuration of the

circuit components in a concrete pool, which was tentatively shown to reduce

the construction costs.

On-site fuel cycle arrangement is also likely to be economically beneficial

owing to the shorter out-of-pile cooling and transportation time, which will

eventually lead to a reduction in the recycled fuel quantity – one of the greater

contributors to the fuel cycle costs.

BREST-OD-300 being a prototype of the prospective commercial BREST plant,

both facilities are quite similar in their design and performance

(V.Lemekhov, V.S. Smirnov, Fast reactor with lead coolant and on-site fuel cycle, Safety of Nuclear Technologies and

Environment, № 1'2012

)

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THE BREST-OD-300 REACTOR DEVELOPMENT

The detailed design of the BREST-OD-300 reactor is being

developed as part of the PRORYV project, including

calculations and experiments conducted to justify the

engineering and process approaches.

Currently the engineering design of the reactor, including

experimental study on small and medium-sized stands

and work stations, as well as on the current study on

verified software tools. Further justification will be held on

the large-scale model.

The main areas of research are: the active zone, the main

process equipment (reactor vessel, reactor coolant pump,

steam generator, etc.), the technology of the coolant, the

study of transitional and emergency processes, including

the closed NFC issues.

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CONCLUSION

1. Project BREST-OD-300 creates a base for development of commercial reactor for

NPP on the basis of new nuclear technologies.

2. Analysis of transient processes in RU BREST-OD-300 shows a possibility of

exclusion of heavy accidents, demanding evacuation and displacement of inhabitants

while using first physical properties of coolant, fuel, other reactor components, аnd

also technical design, directed at it realization.

3. Substantiation of adopted decisions for design project in 2014 is built on

experimental substantiation on small- and middle-scale benches and working areas,

and also on computational substantiation with verified program means. In the further

the substantiation will be held on large-scale mockups.

4. The results of experimental and design works points at possibility of realization in

power complexes with RU of a kind BREST in closed fuel cycle basic demands of

NP on safety, volume of consumption of fuel raw materials, efficiency, solving

problem of spent fuel.

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CONCLUSION

PROBLEMS, PROSPECTS, COLLABORATION:

Nitride fuel:

There are in-pile test of experimental FAs with U-Pu-N fuel in BN-

600 and BOR-60 reactors.

Post-irradiation tests (swelling, gas release, etc.)

Must be justified:

Nitride fuel with MA: MA transmutation mode – homogeneous or

heterogeneous ?

Nuclear Safety for active zones fuelled with nitride fuel for the

heterogeneous transmutation of MA;

•Curium - transmutation or storage?

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CONCLUSION

PROBLEMS, PROSPECTS, COLLABORATION:

•The experimental data are needed on the verification impacts of MA

on neutron-physical and safety characteristics.

•Nuclear and reactor data (measurements, verification, libraries).

• Modeling of MA behavior in radiochemical processes, in fuel,

during long-term storage.

• Risks for proliferation.

• FP transmutation ?

•Approaches and achieving of the Radiation Equivalence

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ASC «Rosatom» company, JSC «NIKIET»

P.O.B 788, 101000, Moscow, Russia

E-mail: nikiet@nikiet.ru

THANK YOU FOR YOUR ATTENTION!