ГОСУДАРСТВЕННАЯ КОРПОРАЦИЯ ПО АТОМНОЙ ЭНЕРГИИ «РОСАТОМ»

Pilot demonstration reactor BREST-OD-300: conceptual

approaches and their implementation

Andrei V. Moiseev (JSC NIKIET, Moscow, Russia)

9th Joint IAEA–GIF Technical Meeting/Workshop

on the Safety of Liquid Metal Cooled Fast Reactors

IAEA Headquarters, Vienna, Austria

WebEx Virtual Meeting

30 March – 01 April 2021

2

Target requirements for commercial reactors with

lead coolant

Elimination of NPPs accidents requiring

population evacuation, much less resettlement

Establishment of CNFC for full

utilization of energy potential of natural

raw uranium – use of VVER Pu for

starting loading

Progressive approximation to

radiation-equivalent (in relation to

natural raw materials) RW disposal –

at the operating stage after

development of fuel with MA

Commercial reactor

Technological enhancement to non-

proliferation regime - no blanket,

no Pu extraction during SNF reprocessing,

on-site NFC, no uranium enrichment required

Marketability against other types of

power generation - demonstration of

technology potential

BREST-OD-300

3

Design concept of BREST-OD-300

MCP Steam generator

Vessel Reactor core

ECCS collector

Mixed nitride fuel with high density and thermal

conductivity, allows to ensure full reproduction of fuel

in the core (core reproduction ratio ~ 1.05) and

compensation of reactivity at fuel burnout.

Lead coolant with high boiling point, low activated,

not entering into violent interaction with water and air

in case of circuit depressurization.

Integral layout in combination with multi-layer metal-

concrete vessel (no coolant escape to beyond the

vessel) to exclude coolant losses.

No shutoff valves in the primary circuit – no

circulation can be lost. A coolant circulation pattern

with a free level difference – circulation is safely

continued during loss of power.

Passive emergency cooling system with natural air

circulation and heat removal to the atmosphere.

Preferential use for safety: neutron-physical

and physico-chemical properties of fuel,

coolant, materials, as well as design solutions

that allow to fully realize these properties.

4

Specific indicators of BREST-OD-300

Thermal power, MW 700

Steam output, t/h, not less than 1500

Maximum neutron flux in the core, cm-2s-1 3.51015

Fuel (U-Pu)N

Fuel loading, t / FAs 20.8 / 169

Burnup at a point (annual refuelings):

- for 6 % (~60 FAs), t

- for 10 % (~35 FAs), t

~7.2

~4.2

Maximum burnup, % h.a. up to 10

Maximum damaging dose at the fuel

cladding at a burnup of 9 % h.а., dpa up to 140

Number of circulation loops 4

Maximum (hydrostatic) pressure of

primary coolant, MPa 1.17

Cover gas pressure (argon) above the

coolant level, MPa:

- during normal operation

- maximum

~ 0.104

0.2

Average mixed temperature of lead

coolant in core inlet/outlet, °С 420/535

Secondary working medium Water

(steam)

Secondary coolant parameters

(water-steam):

- SG inlet pressure, MPa

- SG outlet pressure, MPa

- SG inlet temperature, °С

- SG outlet temperature, °С

18.5

17

340

505

Efficiency, % 43.5

Design life, years, not less than 30

5

Reactor core

Full fuel reproduction in the core – fuel density 12.3 g/cm3;

core reproduction coefficient (no blanket) is ~ 1.05; the ability to work with MA

Small reactivity margin in the power range 30-100% ~ 0,65 βeff (w/o Np-effect)

Average core power density – 175 MWt/m3

Maximum linear power along the core – 420 Wt/cm

Two radial sub-cores for regulation of radial power distribution and the coolant flow

Hexagonal FAs without shrouds:

- cladding temperature no more 800˚

at overlapping even 7 FAs

coolant section

- reduction in FA metal

consumption by more than 30%

- using of manufacturing

experience of VVER FAs

1 – Fuel pin; 2 – Spacer grid; 3 – FA head;

4 – Supporting tube; 5 – FA tail; 6 – Control rods

FA central

FA peripheral

FA with AR rods

FA with RC rods

FA with safety rods

SPF

Reflector

6

Core elements, nitride fuel, materials

FA mockups (all types) and reflector blocks were produced under the industrial

conditions. Manufacturing technology is fully developed.

Mechanical tests carried out for FA mockups and reflector blocks (compression, tension,

simulation of standard and seismic loads). Vibration, strength and vibrometric tests were

performed in the flow of water and lead coolant.

The hydraulic characteristics of fuel assemblies and reflector blocks of all types in the

flow of water and lead were determined. It was shown that hydraulic calculations for

liquid lead can be performed in the same way as for water.

Loading-unloading FA mockups from the core was experimentally tested

Full-scale FA

mock-up

Post-irradiation studies of BREST-type fuel elements with EP823

steel cladding completed. Rector tests were carried out in BOR-60

and BN-600. In total, more than 1000 fuel elements with MNUP fuel

were irradiated.

EFA-11– irradiated in the BN-600 core, achieved parameters:

more than 9 % h.a. and 100 dpa.

All semi-finished products from EP823 steel were put into

production. All properties have been obtained that ensure the

operability of the fuel elements up to 6% h.a. burnout: short-term,

long-term, under irradiation, in a coolant medium.

Pellet of irradiated

MNUP fuel in the

middle section

7

Reactor vessel

Vessel shaft

Pipes of SG leaks

localization

system

Pipelines of the

heating system

Vessel height, m 17,5

Vessel diameter, m 26

Weight including concrete,

thousands of tons 11,8

Service life, not less, years 30

Pipelines of the

cooling system

The vessel of the reactor is made of multilayer reinforced

concrete, lead coolant and the main primary circuit equipment

are located in the vessel.

The temperature on the outer surface of the reactor vessel is

not more than 60°С.

Experimentally determined properties of high-temperature

concretes at temperatures of 400-700° C and under irradiation.

The chemical inertness of the coolant in relation to concrete

was shown.

Estimated partial (not dangerous) loss-of-coolant probability –

less than 9,7x10-10 1/year

Vessel model 1:10

Bottom mockup

8

Steam Generator

Feed water

chamber

Steam

chambers

Shell

Entrance of the

coolant

Mounting

flange

Heat exchanger

tubes

Thermal power of SG unit, MW 90

Height, m 12,1

Diameter of the immersed part, m 2,0

Weight, t 69

Steam productivity of a SG module, kg/s 52,40

Service life, years 30

Monometallic tubes, corrosion resistant in water and

lead, no welds along the entire length. Twisted heat

exchange part.

Internal extended throttles that provide hydrodynamic

stability and limit the steam output in case of

depressurization of the pipe.

Absence of dependant failure in case of one tube rupture

was experimentally demonstrated.

With the passage of steam bubbles through the core, the

maximum realizable reactivity effect Δρmax << 0,5 βeff

X-ray images of jet-flow

(treg = 3 s, treg = 5 s)

The experiment on

demonstration of tube rupture

9

Main circulation pump & Other equipment

MCP

Impeller was optimized using

the medium-scale water and lead test

benches.

Obtained required head-flow curve

ensuring pump operation within

30-100% range.

Endurance tests of the bearing using lead

are conducted (with >70% of the design life

achieved).

Full-scale bench for MCP testing in lead

is being created.

Other equipment

CPS actuator prototype tests was

competed with positive outcome.

Engineering design of the reactor

automated inspection and control

system has been developed; a bench

had been developed, which was used to

demonstrate operation stability CPS

channel regulators during various

transients.

Endurance testing of coolant quality

system components are

conducted.

Testing of the elements of the fuel

element cladding control

system is completed.

Engine

Shaft

Impeller

Discharge

Shaft

cooler

Radial-axial

bearing

Секции

корпуса

Sealing

block

10

Research of output of FP and AP out of the coolant

Two out-of-reactor loop units and one in-reactor loop unit were established

Data on outputs of 210Po, 131I, 115mCd, 110mAg, 123mTe, 210Hg, 124Sb were obtained at temperatures of 500,

610 and 680 °C with lead surface blowing and lead gas bubbling

Experimental data on mass transfer of gaseous (Kr, Xe) and volatile (I) fission products from the nitride

fuel into the gas environment (He) for the code verification were obtained

Processing of the experimental results showed that the main part of the caesium contained in lead (more

than 95%) was deposited on the inner surfaces of the lead circuit of the experimental facility

The requirements for the composition of the initial providing lead

(content of impurities) for the BREST coolant was determined

11

Validation of neutron-physical, thermohydraulic,

and dynamic computation codes

Completed the development and validation of neutron-physical codes MCU, FACT on the previously

performed experiments at the BFS stand (more than 10 critical assemblies) and the BN-350, BN-600

reactors. Neutron-physical experiments with nitride fuel at the BFS were carried out: the deviation of the

experimental values of the criticality parameters from the calculated values is less than 0,2 % Δk/k

For validation of cellular (PUCHOK) and CFD codes experimentally obtained heat transfer coefficients in a

typical fuel pin bundle under conditions of LMC and intercell mixing coefficients in bundles and on the

border of fuel pin bundles

The safety justification of the BREST-OD-300 was performed using the coupled neutron-physical and

thermal-hydraulic code DINAR: validated on experimental data (BOR-60, SPRUT, EUST, etc.). Cross-

verification was performed with the following codes: MCU, EUCLID, HYDRA-IBRAE. Thermohydraulic

“verification” CFD calculations was carried out to justify safety.

Benchmark model of the

BREST-OD-300 core.

The BFS-113 assembly with nitride

Distribution of the cladding

temperature in the core at the overlap

of 7 central FAs

Fuel assembly mock-up for testing

and CFD calculation

12

The greatest differences in the analysis of reactivity

accidents of VVER, BN, BREST reactors

Introducing of the full reactivity margin without triggering the shutdown

systems:

for VVER is not considered, estimates are made

for BN calculations are in progress

for BREST is considered on the basis of 24 hours

Target requirements for BREST-OD-300 :

Elimination of fuel melting (Тfuel melt > 2700 °C) and cladding (Тclad. melt ~ 1500 °C).

Elimination of coolant boiling (Тboil ~ 1750 °C).

Maintaining the integrity of the primary circuit.

The requirement to limit the impact outside the site – excluding the need for

evacuation and resettlement of the population.

13

Introducing of the full reactivity margin at nominal power.

Scenario realization

Scenario:

• initiating event – introducing of the full

reactivity margin at nominal power

0.65eff with realistic design speed (in 30

seconds)

• active protection systems fail (total number

of failures – 11, probability of realization

2.8*10-9)

• pump shutdown on setpoint temperature of

the coolant inlet at SG – 520 С

• as a consequence - loss of forced cooling of

the core and transition to Pb natural

circulation

• passive feedback system (PFS) works due

to the decrease in coolant pressure (input

negative reactivity – 0.63 eff )

0

0,2

0,4

0,6

0,8

1

1,2

1,4

1,6

1,8

2

0 100 200 300 400 500

N; G, отн.ед.

Время, с

N

G

Time, s

Time, s

N; G, rel. units

Tfuel

Tclad

Тcore in

Тcore out

ТSG out

ТSG in

14

Scenario development - shutdown (power reduction)

of the reactor due to:

• temperature equalization of fuel, construction

materials and coolant – increase in average

temperatures and Doppler effect

• temperature reactivity effect due to expansion of the

core with increasing coolant temperature at the core

inlet

Results of calculations of steady-state values:

• maximum fuel temperature 950 - 800 C

• maximum cladding temperature 950 - 800 C

Introducing of the full reactivity margin at nominal power

Steady-state values

Time, s

Time, s

N; G, rel. units

Tfuel

Tclad

Тcore in

Тcore out

ТSG in

ТECCS in

15

Map of maximum temperature distribution of fuel

cladding (at 125 s)

Outcome:

• melting of fuel and cladding, boiling of the coolant -

does not occur

• circulation circuit integrity ensured

• in the part of FAs (distribution - in fig), the maximum

design limit on the cladding temperature is exceeded

• fuel elements may lose their tightness, but they are not

destroyed

• emissions into the atmosphere for all normalized

radionuclides do not reach the reference level per

day

Nuclide Emission, Bq / day Reference level, Bq / day

Cs-134 1.9*106 2.5*106

Cs-137 3.6*106 5.6*106

I-131 5.2*106 5.0*107

Kr-85 7.3*108

1.9*1012

Xe-133 4.6*1010

Xe-133m 9.6*108

Xe-135 1.3*1010

Sum Kr+Xe 6.1*1010

General conclusion:

There is no need to evacuate and resettle the

population outside the industrial site (for

severe accidents with the probability greater

than 3.2∙10-8 1/year).

Introducing of the full reactivity margin at nominal power.

Outcome

16

BREST-OD-300. Complete electric breakdown of the unit (1/2)

0

0,2

0,4

0,6

0,8

1

1,2

0 50 100 150 200 250 300 350

N; G, rel. units

Time, s

N

G

300

400

500

600

700

800

900

1000

1100

1200

1300

0 50 100 150 200 250 300 350

T, С

Time, s

Tfuel

Tclad

TSG in Тcore out

TSG out

Тcore in

Scenario:

• the greatest deterioration of the conditions of heat

removal from the reactor core - total blackout

• shutting down four MCPs and stopping the supply of

feed water when operating at the nominal power

• removal of residual energy is carried out by two of the

four ECCS loops (postulated failure of two emergency

decay heat removal loops, postulated failure of normal

cooling system)

Outcome:

Maximum cladding temperature of the most loaded fuel

element during ~ 45 s exceed 800 С and achieves ~ 890 С

Melting of fuel rod cladding and fuel does not occur. The

integrity of the circuit is ensured.

FP output from the reactor unit for the first day does not

exceed a reference level of emissions per day during normal

operation.

17

Decay heat from the reactor is diverted by the operation

of only two ECCS loops.

Outcome:

Melting of fuel element cladding and fuel does not occur.

The integrity of the circuit is ensured.

FP output from the reactor unit for the first day does not

exceed a reference level of emissions per day during

normal operation. 300

350

400

450

500

550

600

650

700

750

800

0 5000 10000 15000 20000 25000 30000 35000

T, С

Time, s

Behavior of fuel temperature, fuel cladding,

lead coolant at the inlet and outlet of core

and SG with prolonged cooldown.

BREST-OD-300. Complete electric breakdown of the unit (2/2)

General conclusion:

There is no need to evacuate and resettle the population

outside the industrial site (for severe accidents with the

probability greater than 3.2∙10-8 1/year).

18

Regulatory acts and standardization

The reactor design is based on the requirements of the existing federal norms and regulations for NPPs:

general regulations for safety assurance (OPB), nuclear safety rules (PBYa), radiation safety standards

(NRB), etc.

To provide development of innovative NPPs, the processes of the development of design, new federal

norms and regulations and advanced software tools (computational codes) are almost parallel

Rules for arranging and safe operation and the corresponding documentation (welding, control regulations,

etc.), vessel strength calculations norms are specific

Currently, activities in collaboration with Rostekhnadzor are ongoing on approval and implementation of the

fundamental norms and regulations based on which reactor is developed

New federal norms and

regulations

Set of standards

19

The implementation of the project in the pilot

demonstrational energy complex (PDEC)

Practical confirmation of the main technical solutions used in the lead-cooled reactor operating in

closed NFC

Step-by-step substantiation of the resource characteristics of the reactor elements for the creation of

commercial NPPs with lead-cooled reactors

Generation of thermal energy for the subsequent conversion into electrical energy as part of the

power unit, conducting R&D at PDEC

Construction license for BREST-OD-300 issued by Rostechnadzor on 10 Feb 2021

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Conclusion

1. On the basis of the concept of natural (inherent) safety, technical solutions were developed for the

creation of the BREST-OD-300 reactor.

2. Engineering solutions for the reactor equipment have been confirmed by the positive results of

experiments on the models of equipment or its components, computational justification taking into

account the influence of the lead coolant.

3. The design solutions for the core have been confirmed by positive results of irradiation

experiments, hydraulic and vibration experiments, and calculations using certified codes.

4. Severe accidents analysis confirmed one of the target indicators - there is no need to evacuate

and resettle the population outside the industrial site (for severe accidents with the probability

greater than 3.2∙10-8 1/year).

5. The results obtained during the development of the BREST project formed the basis for the

development regulatory framework for the fast reactors with lead coolant.

6. An expert review of the Russian Academy of Sciences was carried out, which confirmed that the

design of the BREST-OD-300 power unit corresponds to the current level of science and

technology, scientific ideas about the problems of the existing nuclear power plant and ways to

solve them. The expertise recommended the construction of the power unit.

7. The BREST-OD-300 received construction license at the beginning of 2021, the construction may

start soon.

8. BREST-OD-300 is an experimental power unit, on which R&D should be completed. Many R&D

results cannot be obtained without creating this reactor.

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Thank you for your attention!