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)
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
2
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
2
3
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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4
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/
6
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
7
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
12
The most severe failures of normal operation
13
The most severe failures of normal operation
14
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