development of physical conceptions of fast reactors · pdf filedevelopment of physical...
TRANSCRIPT
Development of physical
conceptions of fast reactors
YuS Khomyakov1) VI Matveev2) AV Moiseev2)
1)Institution ITC ldquoPRORYVrdquo Project Moscow Russia
2)State Scientific Center of the Russian Federation ndash Institute for Physics and Power Engineering Obninsk Russia
International Conference on Fast Reactors and Related Fuel Cycles Safe Technologies and Sustainable Scenarios (FR13) Paris France 4-7 March 2013
2
Main stages of fast reactor progress in Russia
1 Creation of fundamental basis of FRs (1950-1970)
critical zero-power BR-1
experimental reactors BR-510 BOR-60
2 Engineering and technical familiarization of Sodium
Fast Reactors (1970-1990)
first prototype of fast reactor BN-350
power fast reactor BN-600 of Beloyarsk NPP (up to 2020)
3 Discussions and conceptual investigations (1990-2010)
4 Current Russian Program (2010-2020)
commercialization SFR and development new type FR
power MOX ndashfuel reactor BN-800 with sodium coolant
commercial reactor BN-1200 with sodium coolant
prototype of LFR fast reactor BREST-OD-300 with Pb coolant
prototype of LVFR type reactor with Pb-Bi coolant
experimental fast reactor ndash MBIR with sodium coolant
3
Breeding idea and laquominimal T2 conceptraquo
BR-1 measured Breeding Ratio
BR = 25 plusmn 02
Basic points of laquominimal T2 conceptraquo
small critical mass high core power density
high breeding
short fuel cycle
Na ndash unique option of coolant
Development Na technology - main task
4
First power experimental fast reactor BR-510
BR-2 (1956-1957)
fuel ndash metallic Pu
power - 100 kW
coolant ndash Mercury (Hg)
BR-5 (1959-1964)
fuel ndash plutonium oxide
power - 5 MW
coolant ndash Sodium (Na)
BR-10 (1964-1998)
fuel ndash monocarbide and mononitride U
power - 10 MW
coolant ndash Sodium (Na)
neutron flux up to 151015 cm-2s-1
BR-10 experience oxide and nitride fuel
5
First prototype fast reactor BN-350
The BN-350 reactor (1000 MW(th) 350 MW(e))
was the Worldrsquos first fast reactor-prototype the
loop-type reactor was cooled through 6
separate loops with sodium coolant
BN-350 confirm technical and engineering
reliability of fast reactors and gave first real
experience of operation
Problems with reliability of steam and gas
generators ndash the critical moment for BN program
Na coolant technology ndash the main problem of
fast reactors development
6
BN-600 reactor ldquoNa technology can be saferdquo
BN-600 has finally defined laquo classical imageraquo Russian BN
reactor
bull integral type of layout
bull three circulation loops (radioactive Na-Na heat
exchangers and nonradioactive Na-H2O steam generators)
bull oxide fuel (UO2 in BN-350 and in BN-600)
bull three zones of enrichment of core
bull availability of radial and axial blankets
Classical parameters of a core
bull power density ~ 450 kWm3 and ~ 48 kWm
bull burnup ~ 10-11 hm damage of cladding -80-90 dpa
bull output temperature of Na ~550С steel ~700С
bull operation time between fuel reloading ndash frac12 year
Basic trends of modern development of fast reactors
physical conceptions
safety achievement of core inherent safety
economy improvement of economical characteristics
U238 resources assimilation of closed nuclear fuel cycle
radioactive wastes transmutation of minor actinides
non-proliferation prevention of weapon-grade Pu
production and extraction of pure Pu
7
Key problems and ldquochoice forksrdquo
1 evaluation of feasibility of inherently safe fast reactors
2 choice of coolant sodium heavy liquid metal gas or steam
3 choice of fuel type МОХ carbide nitride or metal
4 expediency of use of fertile blankets
5 expediency and method (hetero- homo-geneous) of MA transmutation
6 fuel breeding level from BR~1 to BR~15
7 fuel breeding level in the core core with equilibrium fuel and BRcore ~1
8 power density in the core from ~500 MWm3 to ~250 MWm3 and lower
9 optimal fuel burn-up value from ~10 to ~15 and to ~20
10 fuel cycle duration from 1-3 years to 5 years and more
11 depth of fuel purification in reprocessing from 10-4 to 10-8
8
9
Breeding trend from BR-1 to BN-800
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening
oxide fuel (light element - O)
sodium coolant ( - Na)
priority of safety
Na plenum and
exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
2
Main stages of fast reactor progress in Russia
1 Creation of fundamental basis of FRs (1950-1970)
critical zero-power BR-1
experimental reactors BR-510 BOR-60
2 Engineering and technical familiarization of Sodium
Fast Reactors (1970-1990)
first prototype of fast reactor BN-350
power fast reactor BN-600 of Beloyarsk NPP (up to 2020)
3 Discussions and conceptual investigations (1990-2010)
4 Current Russian Program (2010-2020)
commercialization SFR and development new type FR
power MOX ndashfuel reactor BN-800 with sodium coolant
commercial reactor BN-1200 with sodium coolant
prototype of LFR fast reactor BREST-OD-300 with Pb coolant
prototype of LVFR type reactor with Pb-Bi coolant
experimental fast reactor ndash MBIR with sodium coolant
3
Breeding idea and laquominimal T2 conceptraquo
BR-1 measured Breeding Ratio
BR = 25 plusmn 02
Basic points of laquominimal T2 conceptraquo
small critical mass high core power density
high breeding
short fuel cycle
Na ndash unique option of coolant
Development Na technology - main task
4
First power experimental fast reactor BR-510
BR-2 (1956-1957)
fuel ndash metallic Pu
power - 100 kW
coolant ndash Mercury (Hg)
BR-5 (1959-1964)
fuel ndash plutonium oxide
power - 5 MW
coolant ndash Sodium (Na)
BR-10 (1964-1998)
fuel ndash monocarbide and mononitride U
power - 10 MW
coolant ndash Sodium (Na)
neutron flux up to 151015 cm-2s-1
BR-10 experience oxide and nitride fuel
5
First prototype fast reactor BN-350
The BN-350 reactor (1000 MW(th) 350 MW(e))
was the Worldrsquos first fast reactor-prototype the
loop-type reactor was cooled through 6
separate loops with sodium coolant
BN-350 confirm technical and engineering
reliability of fast reactors and gave first real
experience of operation
Problems with reliability of steam and gas
generators ndash the critical moment for BN program
Na coolant technology ndash the main problem of
fast reactors development
6
BN-600 reactor ldquoNa technology can be saferdquo
BN-600 has finally defined laquo classical imageraquo Russian BN
reactor
bull integral type of layout
bull three circulation loops (radioactive Na-Na heat
exchangers and nonradioactive Na-H2O steam generators)
bull oxide fuel (UO2 in BN-350 and in BN-600)
bull three zones of enrichment of core
bull availability of radial and axial blankets
Classical parameters of a core
bull power density ~ 450 kWm3 and ~ 48 kWm
bull burnup ~ 10-11 hm damage of cladding -80-90 dpa
bull output temperature of Na ~550С steel ~700С
bull operation time between fuel reloading ndash frac12 year
Basic trends of modern development of fast reactors
physical conceptions
safety achievement of core inherent safety
economy improvement of economical characteristics
U238 resources assimilation of closed nuclear fuel cycle
radioactive wastes transmutation of minor actinides
non-proliferation prevention of weapon-grade Pu
production and extraction of pure Pu
7
Key problems and ldquochoice forksrdquo
1 evaluation of feasibility of inherently safe fast reactors
2 choice of coolant sodium heavy liquid metal gas or steam
3 choice of fuel type МОХ carbide nitride or metal
4 expediency of use of fertile blankets
5 expediency and method (hetero- homo-geneous) of MA transmutation
6 fuel breeding level from BR~1 to BR~15
7 fuel breeding level in the core core with equilibrium fuel and BRcore ~1
8 power density in the core from ~500 MWm3 to ~250 MWm3 and lower
9 optimal fuel burn-up value from ~10 to ~15 and to ~20
10 fuel cycle duration from 1-3 years to 5 years and more
11 depth of fuel purification in reprocessing from 10-4 to 10-8
8
9
Breeding trend from BR-1 to BN-800
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening
oxide fuel (light element - O)
sodium coolant ( - Na)
priority of safety
Na plenum and
exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
3
Breeding idea and laquominimal T2 conceptraquo
BR-1 measured Breeding Ratio
BR = 25 plusmn 02
Basic points of laquominimal T2 conceptraquo
small critical mass high core power density
high breeding
short fuel cycle
Na ndash unique option of coolant
Development Na technology - main task
4
First power experimental fast reactor BR-510
BR-2 (1956-1957)
fuel ndash metallic Pu
power - 100 kW
coolant ndash Mercury (Hg)
BR-5 (1959-1964)
fuel ndash plutonium oxide
power - 5 MW
coolant ndash Sodium (Na)
BR-10 (1964-1998)
fuel ndash monocarbide and mononitride U
power - 10 MW
coolant ndash Sodium (Na)
neutron flux up to 151015 cm-2s-1
BR-10 experience oxide and nitride fuel
5
First prototype fast reactor BN-350
The BN-350 reactor (1000 MW(th) 350 MW(e))
was the Worldrsquos first fast reactor-prototype the
loop-type reactor was cooled through 6
separate loops with sodium coolant
BN-350 confirm technical and engineering
reliability of fast reactors and gave first real
experience of operation
Problems with reliability of steam and gas
generators ndash the critical moment for BN program
Na coolant technology ndash the main problem of
fast reactors development
6
BN-600 reactor ldquoNa technology can be saferdquo
BN-600 has finally defined laquo classical imageraquo Russian BN
reactor
bull integral type of layout
bull three circulation loops (radioactive Na-Na heat
exchangers and nonradioactive Na-H2O steam generators)
bull oxide fuel (UO2 in BN-350 and in BN-600)
bull three zones of enrichment of core
bull availability of radial and axial blankets
Classical parameters of a core
bull power density ~ 450 kWm3 and ~ 48 kWm
bull burnup ~ 10-11 hm damage of cladding -80-90 dpa
bull output temperature of Na ~550С steel ~700С
bull operation time between fuel reloading ndash frac12 year
Basic trends of modern development of fast reactors
physical conceptions
safety achievement of core inherent safety
economy improvement of economical characteristics
U238 resources assimilation of closed nuclear fuel cycle
radioactive wastes transmutation of minor actinides
non-proliferation prevention of weapon-grade Pu
production and extraction of pure Pu
7
Key problems and ldquochoice forksrdquo
1 evaluation of feasibility of inherently safe fast reactors
2 choice of coolant sodium heavy liquid metal gas or steam
3 choice of fuel type МОХ carbide nitride or metal
4 expediency of use of fertile blankets
5 expediency and method (hetero- homo-geneous) of MA transmutation
6 fuel breeding level from BR~1 to BR~15
7 fuel breeding level in the core core with equilibrium fuel and BRcore ~1
8 power density in the core from ~500 MWm3 to ~250 MWm3 and lower
9 optimal fuel burn-up value from ~10 to ~15 and to ~20
10 fuel cycle duration from 1-3 years to 5 years and more
11 depth of fuel purification in reprocessing from 10-4 to 10-8
8
9
Breeding trend from BR-1 to BN-800
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening
oxide fuel (light element - O)
sodium coolant ( - Na)
priority of safety
Na plenum and
exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
4
First power experimental fast reactor BR-510
BR-2 (1956-1957)
fuel ndash metallic Pu
power - 100 kW
coolant ndash Mercury (Hg)
BR-5 (1959-1964)
fuel ndash plutonium oxide
power - 5 MW
coolant ndash Sodium (Na)
BR-10 (1964-1998)
fuel ndash monocarbide and mononitride U
power - 10 MW
coolant ndash Sodium (Na)
neutron flux up to 151015 cm-2s-1
BR-10 experience oxide and nitride fuel
5
First prototype fast reactor BN-350
The BN-350 reactor (1000 MW(th) 350 MW(e))
was the Worldrsquos first fast reactor-prototype the
loop-type reactor was cooled through 6
separate loops with sodium coolant
BN-350 confirm technical and engineering
reliability of fast reactors and gave first real
experience of operation
Problems with reliability of steam and gas
generators ndash the critical moment for BN program
Na coolant technology ndash the main problem of
fast reactors development
6
BN-600 reactor ldquoNa technology can be saferdquo
BN-600 has finally defined laquo classical imageraquo Russian BN
reactor
bull integral type of layout
bull three circulation loops (radioactive Na-Na heat
exchangers and nonradioactive Na-H2O steam generators)
bull oxide fuel (UO2 in BN-350 and in BN-600)
bull three zones of enrichment of core
bull availability of radial and axial blankets
Classical parameters of a core
bull power density ~ 450 kWm3 and ~ 48 kWm
bull burnup ~ 10-11 hm damage of cladding -80-90 dpa
bull output temperature of Na ~550С steel ~700С
bull operation time between fuel reloading ndash frac12 year
Basic trends of modern development of fast reactors
physical conceptions
safety achievement of core inherent safety
economy improvement of economical characteristics
U238 resources assimilation of closed nuclear fuel cycle
radioactive wastes transmutation of minor actinides
non-proliferation prevention of weapon-grade Pu
production and extraction of pure Pu
7
Key problems and ldquochoice forksrdquo
1 evaluation of feasibility of inherently safe fast reactors
2 choice of coolant sodium heavy liquid metal gas or steam
3 choice of fuel type МОХ carbide nitride or metal
4 expediency of use of fertile blankets
5 expediency and method (hetero- homo-geneous) of MA transmutation
6 fuel breeding level from BR~1 to BR~15
7 fuel breeding level in the core core with equilibrium fuel and BRcore ~1
8 power density in the core from ~500 MWm3 to ~250 MWm3 and lower
9 optimal fuel burn-up value from ~10 to ~15 and to ~20
10 fuel cycle duration from 1-3 years to 5 years and more
11 depth of fuel purification in reprocessing from 10-4 to 10-8
8
9
Breeding trend from BR-1 to BN-800
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening
oxide fuel (light element - O)
sodium coolant ( - Na)
priority of safety
Na plenum and
exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
5
First prototype fast reactor BN-350
The BN-350 reactor (1000 MW(th) 350 MW(e))
was the Worldrsquos first fast reactor-prototype the
loop-type reactor was cooled through 6
separate loops with sodium coolant
BN-350 confirm technical and engineering
reliability of fast reactors and gave first real
experience of operation
Problems with reliability of steam and gas
generators ndash the critical moment for BN program
Na coolant technology ndash the main problem of
fast reactors development
6
BN-600 reactor ldquoNa technology can be saferdquo
BN-600 has finally defined laquo classical imageraquo Russian BN
reactor
bull integral type of layout
bull three circulation loops (radioactive Na-Na heat
exchangers and nonradioactive Na-H2O steam generators)
bull oxide fuel (UO2 in BN-350 and in BN-600)
bull three zones of enrichment of core
bull availability of radial and axial blankets
Classical parameters of a core
bull power density ~ 450 kWm3 and ~ 48 kWm
bull burnup ~ 10-11 hm damage of cladding -80-90 dpa
bull output temperature of Na ~550С steel ~700С
bull operation time between fuel reloading ndash frac12 year
Basic trends of modern development of fast reactors
physical conceptions
safety achievement of core inherent safety
economy improvement of economical characteristics
U238 resources assimilation of closed nuclear fuel cycle
radioactive wastes transmutation of minor actinides
non-proliferation prevention of weapon-grade Pu
production and extraction of pure Pu
7
Key problems and ldquochoice forksrdquo
1 evaluation of feasibility of inherently safe fast reactors
2 choice of coolant sodium heavy liquid metal gas or steam
3 choice of fuel type МОХ carbide nitride or metal
4 expediency of use of fertile blankets
5 expediency and method (hetero- homo-geneous) of MA transmutation
6 fuel breeding level from BR~1 to BR~15
7 fuel breeding level in the core core with equilibrium fuel and BRcore ~1
8 power density in the core from ~500 MWm3 to ~250 MWm3 and lower
9 optimal fuel burn-up value from ~10 to ~15 and to ~20
10 fuel cycle duration from 1-3 years to 5 years and more
11 depth of fuel purification in reprocessing from 10-4 to 10-8
8
9
Breeding trend from BR-1 to BN-800
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening
oxide fuel (light element - O)
sodium coolant ( - Na)
priority of safety
Na plenum and
exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
6
BN-600 reactor ldquoNa technology can be saferdquo
BN-600 has finally defined laquo classical imageraquo Russian BN
reactor
bull integral type of layout
bull three circulation loops (radioactive Na-Na heat
exchangers and nonradioactive Na-H2O steam generators)
bull oxide fuel (UO2 in BN-350 and in BN-600)
bull three zones of enrichment of core
bull availability of radial and axial blankets
Classical parameters of a core
bull power density ~ 450 kWm3 and ~ 48 kWm
bull burnup ~ 10-11 hm damage of cladding -80-90 dpa
bull output temperature of Na ~550С steel ~700С
bull operation time between fuel reloading ndash frac12 year
Basic trends of modern development of fast reactors
physical conceptions
safety achievement of core inherent safety
economy improvement of economical characteristics
U238 resources assimilation of closed nuclear fuel cycle
radioactive wastes transmutation of minor actinides
non-proliferation prevention of weapon-grade Pu
production and extraction of pure Pu
7
Key problems and ldquochoice forksrdquo
1 evaluation of feasibility of inherently safe fast reactors
2 choice of coolant sodium heavy liquid metal gas or steam
3 choice of fuel type МОХ carbide nitride or metal
4 expediency of use of fertile blankets
5 expediency and method (hetero- homo-geneous) of MA transmutation
6 fuel breeding level from BR~1 to BR~15
7 fuel breeding level in the core core with equilibrium fuel and BRcore ~1
8 power density in the core from ~500 MWm3 to ~250 MWm3 and lower
9 optimal fuel burn-up value from ~10 to ~15 and to ~20
10 fuel cycle duration from 1-3 years to 5 years and more
11 depth of fuel purification in reprocessing from 10-4 to 10-8
8
9
Breeding trend from BR-1 to BN-800
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening
oxide fuel (light element - O)
sodium coolant ( - Na)
priority of safety
Na plenum and
exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
Basic trends of modern development of fast reactors
physical conceptions
safety achievement of core inherent safety
economy improvement of economical characteristics
U238 resources assimilation of closed nuclear fuel cycle
radioactive wastes transmutation of minor actinides
non-proliferation prevention of weapon-grade Pu
production and extraction of pure Pu
7
Key problems and ldquochoice forksrdquo
1 evaluation of feasibility of inherently safe fast reactors
2 choice of coolant sodium heavy liquid metal gas or steam
3 choice of fuel type МОХ carbide nitride or metal
4 expediency of use of fertile blankets
5 expediency and method (hetero- homo-geneous) of MA transmutation
6 fuel breeding level from BR~1 to BR~15
7 fuel breeding level in the core core with equilibrium fuel and BRcore ~1
8 power density in the core from ~500 MWm3 to ~250 MWm3 and lower
9 optimal fuel burn-up value from ~10 to ~15 and to ~20
10 fuel cycle duration from 1-3 years to 5 years and more
11 depth of fuel purification in reprocessing from 10-4 to 10-8
8
9
Breeding trend from BR-1 to BN-800
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening
oxide fuel (light element - O)
sodium coolant ( - Na)
priority of safety
Na plenum and
exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
Key problems and ldquochoice forksrdquo
1 evaluation of feasibility of inherently safe fast reactors
2 choice of coolant sodium heavy liquid metal gas or steam
3 choice of fuel type МОХ carbide nitride or metal
4 expediency of use of fertile blankets
5 expediency and method (hetero- homo-geneous) of MA transmutation
6 fuel breeding level from BR~1 to BR~15
7 fuel breeding level in the core core with equilibrium fuel and BRcore ~1
8 power density in the core from ~500 MWm3 to ~250 MWm3 and lower
9 optimal fuel burn-up value from ~10 to ~15 and to ~20
10 fuel cycle duration from 1-3 years to 5 years and more
11 depth of fuel purification in reprocessing from 10-4 to 10-8
8
9
Breeding trend from BR-1 to BN-800
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening
oxide fuel (light element - O)
sodium coolant ( - Na)
priority of safety
Na plenum and
exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
9
Breeding trend from BR-1 to BN-800
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening
oxide fuel (light element - O)
sodium coolant ( - Na)
priority of safety
Na plenum and
exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
Modern requirements to the closed fuel cycle
Minimal T2 is not highest priority
bull significant amount of Pu from VVER and
RBMK reactors
bull expected medium rate new nuclear
power
Fuel breeding
BR ~ 105 divide 12
Start-up loading (core power density)
mcritical ~ 6 t GW ( qv~ 200 MWm3 )
Duration of external fuel cycle
T le 3 years
10
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
Conception of ldquofast reactor start from U-235rdquo
11
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U
20
70
120
170
220
270
2010 2020 2030 2040 2050 2060 2070 2080 2090 2100
Pow
er
GG
W
year
ldquoTraditionalrdquo scenario
ldquoU-235 startrdquo scenario
Natural U
FP
Fast reactor
Isolation
Fission
Product
U Pu МА Enrichment U Thermal reactor U Pu МА
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
Main conceptual ways of reactor safety improvement
Minimization of excess reactivity for the fuel burn-up
Decrease of sodium void reactivity effect
Use of passive devices for reactivity control
Use of passive devices for decay heat removal
12
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
Conception of ldquozero excess reactivityrdquo
The concept sets as the purpose exception of
reactor runaway on prompt neutrons by limitation of
excess reactivity ρburn-uplt βeff
It means practical refusal of management of
campaign of fuel due to active systems of reactivity
control and
Transition to management due to internal
properties of a core and fuel
ldquoEquilibrium fuelrdquo is the fuel which is not changing
reactivity during burning out
Nitride fuel is preferable
Key issue is maintenance of sufficient accuracy of
forecasting of campaign (BN-600)
13
-300
-250
-200
-150
-100
-050
000
050
0 100 200 300 td
d
kk
BN-1200-Nitr
BN-1200-MOX
BN-800
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
bull Case 1 (reference core) - Power
gradually decreases due to negative
Net reactivity core is heating up and
after 12 minute sodium boiling starts
in the core voiding of the upper core
part gives additional negative
contribution in Net reactivity and
power continues to go down
Reactor self-protection is provided
bull Case 2 (without Na plenum) and 3
(increased height of core) - sodium
boiling results in positive
contribution to Net reactivity reactor
runaway occurs leading to the core
disruption after 20-28 seconds
Reactor self-protection isrsquot provided
Conception of ldquozero sodium void reactivity effectrdquo
0 20 40 60 80 100Time s
00
02
04
06
08
10
Rel
u
nit
s
Power
Relative primary
sodium flow rate
0 5 10 15 20 25Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 10 20 30Time s
00
04
08
12
16
20
Rel
u
nit
s
Power
Relative primarysodium flow rate
0 20 40 60 80 100Time s
-00025
-0002
-00015
-0001
-00005
0
00005
dk
k
0 5 10 15 20 25Time s
-00125
-001
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
0 10 20 30Time s
-00075
-0005
-00025
0
00025
0005
00075
001
00125
dk
k
Reactivity effects
Doppler effect
Axial fuel expansion
SVRE
Net
Behavior of reactor parameters ULOF accident
Case 1
Case 2
Case 3
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
15
Reactivity effects Reactor power and
primary flow rate
0 100 200 300 400 500
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
0 100 200 300 400 500Время с
-00008
-00006
-00004
-00002
0
00002
dkk
MOX
Nitride
0 10 20 30 40 50 60 70 80 90 100
Время с
00
02
04
06
08
10
Отн
ед
Мощность
Расход 1 контур
Due to Doppler effect net
reactivity and power goes down
rapidly
Sodium boiling does not occurs
0 10 20 30 40 50 60 70 80 90 100Время с
-0003
-0002
-0001
0
0001
dkk
Self-protection features of cores with MOX and Nitride fuel
Comments
Power gradually decreases due
to negative Net reactivity core is
heating up and after frac12 minute
sodium boiling starts in the core
voiding of upper core part gives
additional negative contribution
in Net reactivity and power
continues to go down
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
Transmutation BN-350 experience and conclusions
16
101
102
101
102
174
102
U-235 10
Cm(52)
Cm(43)
Am-241
Pu-240
U-233
42 Central plane of reactor Core height Core height
U-235 (13)
15 Ir
Cell 243
244
226 225
242
260 259
Effectiveness of МА transformation into fission
products in power reactors with low power density is
not high
Main reason - parasitic neutron capture
For example ~23 reactions leads to formation of
secondary actinides in BN-350
Secondary activity can exceed initial MA activity
In what sense of transmutation
Nuclides Burn-up of basic
isotope
Accumulation of
secondary actinides
Actinides
burn-up
Am241 35 255 95
Np237 35 25 10
Cm244 44 33 10
Pu240 20 9 11
Pu238 34 12 22
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
Transmutation the general conceptual approach
17
MA transformation into the fuel isotopes
(similarly to the fuel breeding) - the main
method of their transmutation
Irradiation of short lived curium isotopes does
not make sense and therefore these should be
separated and stored till their decay to
plutonium
It is inexpedient to carry out MA
transmutation separately from the basic fuel
circulating in CNFC
Homogeneous transmutation along with fuel
is considered as the most effective approach
Although heterogeneous transmutation is not
excluded however it is effective if only MA
recycling is synchronized with the fuel recycling
242Am 1602 ч
241Am 4322 г
241Pu 144 г
242Pu
376+5 л
242mAm 141 г
242Сm 1628 д
240Pu 6537 л
239Pu 24065 л
238Pu 8774 г
(nγ)
(nγ)
(1-ω)middot( nγ)
83 17
ИП
243Сm 285 г
244Сm 181 г
243Am 7380 л
244Am 101 ч
(n2n)
(nγ)
(nγ)
ω(nγ)
(nγ)
(nγ) (nγ)
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
BN-800 fast reactor
BN-800 is first post-Soviet fast reactor in
Russia (start up at 2014)
o This project is aimed at the development of
the fuel cycle infrastructure and mastering
of the new types of fuel (MOX fuel)
o Sodium plenum making it possible to
assure zero void reactivity effect and
passive safety systems are special features
of BN-800 reactor design
Tests of these elements would lead to the
progress in the area of fast reactor safety
The problem of the proof of
economic efficiency before the
given project is not put
18
1 - vessel 2 -guard vessel 3 - core 4 - core diagrid
5 - core catcher 6 - silo 7 - main sodium pump
8 - upper stationary shielding 9 - large rotating plug
10 ndashcentral rotating plug 11 - protection cap
12 - refueling mechanism 13 - small rotating plug
14 ndash intermediate heat exchanger
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
BN-1200 fast reactor
Fuel cycle
bull fuel ndash mixed oxide or nitride
bull low power density in the core
bull external fuel cycle duration - 3 years
bull BR ndash 12 (oxide) -13 (nitride BRcore ~1)
bull MA utilization in the basic fuel
Safety
bull 2 types of passive control rods
bull flattened core sodium plenum
bull integration of all primary sodium systems in the
reactor vessel to eliminate radioactive sodium
leaks
Economical characteristics
bull optimization of layout approaches
bull increase of load factor by transition to one-year
refuelling interval
bull increase of the fuel burn-up
19
1 - intermediate heat exchanger 2 - reactor vessel
3 - guard vessel 4 - silo 5 - core diagrid
6 - core catcher 7 - reactor core 8 ndash pump nozzle
9 ndash main sodium pump 10 ndash cold trap
11 ndash control rod drives 12 ndash rotating plug
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
Demonstration BREST-OD-300 lead cooled reactor
bull coolant - lead
bull fuel ndash mixed nitride
bull low power density of the core
bull duration of external fuel cycle ndash 1-2 years
bull BR = BRcore ~ 105 without blankets
bull excess reactivity for the fuel burn-up leβeff
bull elimination of the most severe accidents by
means of natural properties specific for the
reactor its fuel coolant as well as reactor
design facilitating implementation of these
properties
bull rough cleaning of fuel from fission products and
curium
bull potential possibility of using enriched uranium
nitride as starting fuel with further transition to
mixed fuel is under investigation
20
1 ndash reactor vessel 2 ndash steam-water collectors
3 ndash control rod drives 4 ndash rotating plug
5 ndash channels of system emergency cooling
6 - main pump 7 ndash reactor core 8 ndash core diagrid
9 ndash steam generator
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
21 21
Conclusion (1)
Russian experience in developing fast reactors has proved clearly scientific
justification of conceptual physical principles and their technical feasibility
However the potential of fast reactors caused by their physical features has
not been fully realized
In order to assure the real possibility of transition to the nuclear power with
fast reactors by about 2030 it is necessary to consistently update fast reactor
designs for solving the following key problems
bull increasing of self-protection level of reactor core
bull improvement of technical and economical characteristics
bull solution of the problems related to the fuel supply of nuclear power and
assimilation of closed nuclear fuel cycle
bull disposal of long lived radioactive waste and transmutation of minor
actinides
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept
22 22
Conclusion (2)
Russian program (2010-2020) on the development of basic concepts of the new
generation reactors implies successive solution of the above problems
New technical decisions will be demonstrated by development and assimilation
of the new reactors
BN-800 ndash development of the fuel cycle infrastructure and mastering of the
new types of fuel
BN-1200 reactor ndash demonstration economical efficiency of fast reactor and
new level of safety
BREST development and demonstration new heavy liquid metal coolant
technology and alternative design concept