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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
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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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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/
5
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
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
8
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:
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:
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
17
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
24
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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