Experimental possibilities of research fast reactor BOR-60

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Experimental possibilities of research fast
reactor BOR-60

• Efimov V.N., Zhemkov I.Yu., Korolkov A.S.
• FEDERAL STATE UNITARY ENTERPRISE STATE SCIENTIFIC
  CENTER
• OF RUSSIAN FEDERATION RESEARCH INSTITUTE OF
  ATOMIC REACTORS

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• Research fast reactor BOR-60 is one of the
  leading experimental facilities of the country
  and of the world intended for testing of a
  variety of fuel, absorbing and structural
  materials that are offered for creation of
  advanced fast, pressurized water, gas-cooled and
  fusion reactors and serving for substantiation of
  the VVER and BN-type reactor service life
  extension. The reactor has been in effective and
  reliable operation for more than 35 years already
  and at present it is practically the only
  research fast reactor that, apart from well
  equipped material science laboratories and
  pilot-scale production engaged in fuel
  fabrication and reprocessing, has unique
  experimental possibilities for complex
  investigation activities in different research
  lines.

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Table 1 Some physical characteristics of the
reactor
Characteristic Value
Reactor heat power, MW 60
Inlet temperature of coolant, ?? 310-330
Outlet temperature of coolant, ?? 530
Fuel UO2 or UO2-PuO2
235U enrichment, 45-90
Maximum Pu concentration, 40
Maximum volumetric power in the core, kW/l 1100
Maximum neutron flux density, cm-2s-1 3.71015
Average neutron energy, MeV 0.4
Neutron fluence per 1 year, cm-2 31022
Damage dose accumulation rate, dpa/y Up to 25
Fuel burnup rate, /y Up to 6
Power non-uniformity factors Axial Radial 1.14 1.15
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Fig. 1. Simplified schematic diagram of the
BOR-60 reactor facility
1 - reactor 2 - intermediate heat exchanger 3
- circulating pump of the first circuit 4 -
steam generator 5 - sodium-air heat
exchanger 6 - circulating pump of the second
circuit 7 - blow fan 8 - turbine 9 - turbine
condenser 10 - deaerator 11 - condensate
pumps 12 - feed pumps 13 - low pressure
heater 14-high pressure heater
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Fig. 2. The BOR-60 reactor section
1 inlet branch pipe, 2 high pressure
chamber, 3 basket, 4 thermal and neutron
reactor vessel shielding, 5 protective
casing, 6 support flange, 7 refueling
channel, 8 driving mechanism of the control and
safety rods, 9 support flange, 10 large
rotating plug, 11 small rotating plug, 12
core and reflector assemblies
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Fig. 3. Pressure plenum

• 1 pressure plenum chamber
• 2 throttle plug
• 3 adjustable plug
• 4 inlet chamber
• 5 inlet chamber bottom
• 6 - throttle
• 7 - throttle
• 8 - throttle
• 9 - gasket
• 10 shell with displacers
• 11 - displacer

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Fig. 4. Cartogram of the BOR-60 reactor
Reactor loading possibility
Cells quantity for S/A for absorbing rods instrumented cells 265 156 7 3
State S/A quantity 85-124
Maximum quantity of the experimental non-fuel S/A in the core 12
Maximum quantity of the experimental fuel S/A in the core 156
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Fig. 5. Radial distribution of average neutron
energy (En), integral energy (Fn) and neutron
flux density with ?gt0.1 Mev (Fn(0.1))
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Fig. 6. Neutron spectrum of the BOR-60 reactor
core - layer (cell number)
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Fig. 7. Neutron spectrum of the BOR-60 reactor
reflector layer (cell number)
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• - For instrumented irradiation a special
  thermometric channel is used allowing allocating
  experimental devices directly in the core (D23).
  The lower part of the experimental device looks
  like a standard S/A (a fixture and a hexagonal
  tube of 44 mm of across flats dimension).
• - In two cells (?43 and D35) it is possible to
  display limited data (thermocouples, neutron
  sensors, etc.).
• - Peripheral cell G01 of the reflector is
  shielded by three assemblies with zirconium
  hydride that allowed mitigating the cell neutron
  spectrum and using it for radioisotope production
  and other purposes.
• - The reactor is equipped with a horizontal (HEC)
  and 9 vertical (VEC) channels outside of the
  reactor vessel. The channels are used mainly for
  irradiation of electro technical materials and
  silicon radiation doping. By the results of the
  HEC neutron physical characteristics study it was
  concluded that the channel can be used for
  medical investigations.

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Table 2Testing conditions of materials and
products in cell D-23
Parameter Value
Neutron flux density, sm-2s-1 21015
Specific radiation energy release in structural materials (with atomic number Z  26?30), W/g 4
Absorbed gamma-radiation dose rate, Gy/s 4.5103
Coefficient of non-uniform radiation density distribution along the core height (450 mm) for neutrons for gamma-radiation 1.13 1.25
Sodium flow rate, m3/h when fed from high pressure chamber when fed from low pressure chamber up to 8 up to 2
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Table 3 Neutron-physical characteristics of the
BOR-60 instrumented cells (Wreactor55 MW)
Cell, row Cell, row ?31, 1 ?43, 3 D23, 5 D35, 8
Radius of the cell center location against the core center, mm Radius of the cell center location against the core center, mm 45 135 196 360
Neutron flux density, 1015 sm-2s-1 Egt0.0 MeV (F0) Egt0.1 MeV (F0.1) Neutron flux density, 1015 sm-2s-1 Egt0.0 MeV (F0) Egt0.1 MeV (F0.1) 3.4 2.8 3.1 2.5 2.5 2.0 1.2 0.6
Damage accumulation rate in steel (DPA), 10-6 d.p.a./s Damage accumulation rate in steel (DPA), 10-6 d.p.a./s 1.4 1.3 1.0 0.2
Kz(AP), relative unit F0 1.15 1.16 1.15 1.12
Kz(AP), relative unit F0.1 1.17 1.17 1.17 1.15
Kz(AP), relative unit DPA 1.18 1.18 1.18 1.16
Kr(CCP), relative unit F0 1.00 1.05 1.09 1.13
Kr(CCP), relative unit DPA 1.01 1.06 1.11 1.31
Neutron flux density fraction with energy exceeding 0.1 MeV, relative unit Neutron flux density fraction with energy exceeding 0.1 MeV, relative unit 0.83 0.82 0.80 0.50
Average neutron energy, keV Average neutron energy, keV 350 320 250 40
Neutron fluence, 1022 sm-2 Egt0.0 MeV 5.5 5.0 4.1 1.9
Neutron fluence, 1022 sm-2 Egt0.1 MeV 4.6 4.1 3.3 1.0
Steel damage dose, d.p.a. Steel damage dose, d.p.a. 24 21 17 4
1 year of irradiation - WT 250 000 MWh, Kz and
Kr axial and radial non-uniformity coefficient.
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Fig. 8. Neutron spectrum of cell G01 of the
BOR-60 reflector
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Fig. 9. BOR-60 HEC and VEC location scheme
1 - HEC, 2 - sand, 3 - oxide, 4 disperser
drive, 5 cast iron, 6 - graphite, 7 -
concrete, 8 - VEC
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Fig. 10. Neutron spectrum of the BOR-60 VEC
(calculations were made on the basis of MMK and
OKS-ROZ-6 programs)
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Fig.11. Neutron spectrum at the HEC inlet and
outlet
Table 4 Neutron flux and gamma-quantum density at
the BOR-60 HEC outlet (sm-2s-1)
HEC  Calculated value Calculated value Experiment value
HEC  Fn Fg Fn
Without Pb-screen, ?ngt0 MeV (0.84?1.2)?1010 9.6?108 (2.9?3.4)?108
Without Pb-screen, ?ngt1.2 MeV (6.2?8.6)?107 - (5.7?6.5)?107
With Pb-screen 3.6?109 2.9?106 -
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Fig. 12. Typical diagram of the BOR-60 reactor
operation
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• Long term investigation of neutron physical,
  heat-hydraulic and dynamic reactor
  characteristics allowed detailed study of reactor
  behavior in different operation modes, creating a
  complex of computation programs for reactor
  operation and experiments performance. As a
  result, calculations authenticity increased to
  support experimental programs, reactor operation
  and its safety substantiation. On the basis of
  great experience of reactor characteristics
  investigations and a verified complex of
  computation programs different methods were
  developed that enable high accuracy control of
  operation modes and parameters of materials
  irradiation in the non-instrumented reactor
  cells.

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Table 5 Irradiation parameters errors,
Parameter Measurement Calculation
Intel (outlet) reactor temperature 1,2 -
Reactor power - 2,5
Reactor flow rate 3 -
Neutron flux (fluence) 7 10
Experimental devices (ED) flow rate 2 -
ED power 7 10
Intel ED temperature 1 1,5
Outlet ED temperature 1 1-3
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• For irradiation of a variety of materials and
  products at different operation modes and
  parameters a complex of specialized test devices
  is used. The test devices consist of capsule
  devices, dismountable material science
  assemblies, autonomous instrumented channels,
  special instrumented S/As etc.
• Simple design of the devices and possibility to
  install them practically into any core or
  reflector cell can be considered an undoubted
  advantage of the devices.
• The main task the developer of the test devices
  faces is creation of the required temperature
  modes at the specimens. For this purpose thermal
  insulating clearances, intensive cooling or
  additional heating due to radiation energy
  release or fuel fission are used. Temperature
  stabilization is achieved as a result of
  thermistor change in the scheme of heat transfer
  due to the coolant temperature change or as a
  result of the heat removal intensification by
  using liquid metal under boiling condition. These
  devices help to provide the specified axial and
  azimuthal temperature non-uniformity.

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Fig. 13. Flow experimental assemblies
Fig. 14. Device with evaporativewith gas heat
insulation
thermosiphon
1 specimens 2 shell 3 heater 4 body
1,2 outer and inner bodies 3 clearance 4
shells 5 - specimens
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The lower boundary of the irradiation temperature
range that is ensured in the BOR-60 reactor makes
up 300-310??. It significantly expands the scope
of reactor work, including experiments on
investigation of physical and mechanical
properties of zirconium alloys and materials of
the VVER-type reactor internals. At relatively
high coolant flow rate the dismountable assembly
allows irradiating structural materials specimens
at the temperature close to the reactor inlet
temperature. This assembly is one of the simplest
and widely used experimental devices helping to
perform intermediate reloading procedures and
investigation of specimens with their subsequent
irradiation. The dismountable assembly is also
used for irradiation of fuel elements.
1 - thermometric probe 6 - gas clearance 2 -
detachable head 7 - inner pipe 3 - spacer
tubes 8 - capsule assembly 4 - probe
thermocouples 9 - core center 5 - wrapper 10
- fixture

• Fig. 15. Dismountable assembly with a hot probe
  for irradiation of
• structural materials

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1, 2 leak-tight capsules with different type
specimens in lithium-4 medium 3 inner capsule
cladding from Inconel-type heat-resistant
steel 4 outer capsule cladding from stainless
steel 5 ampoule sodium 6 ampoule
clearance 7 leak-tight wrappers from
Inconel-type heat-resistant steel with
thermocouples

• Fig.16. Cross-section of the experimental device
  with capsules for irradiation of vanadium in
  lithium medium

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• 1 - fixture
• 2 - throttling orifice
• 3 - filter
• 4 - block of tungsten rods
• 5 - gas clearance
• 6 - block of steel rods
• 7 - head

• Fig. 17. Scheme of sodium boiling generator

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1 fuel assembly 2 - nozzle body 3 tube with
sensors 4 flow regulator 5 sodium vapor
filter 6 electric engine

• Fig. 18. Scheme of the instrumented nozzle

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1 sodium vapor catcher 2- level gauges inside
of the channel 3 maximum sodium level in the
channel 4- KGO pipe 5 sodium flow regulator 6
sodium yield from electromagnetic pump 7 MGD
pump 8 fuel assembly body 9 sodium upflow in
the channel 10 sodium down flow in the channel
11- upflow of reactor sodium 12 heat
insulating gas clearance of FA in the channel 13
channel body 14 - neutron sensors 15 inner
wrapper of the channel 16 fuel elements 17
membrane 18 tube for sodium channel filling 19
throttling orifice 20 channel tail 21 inlet
of reactor sodium into the channel from the
BOR-60 high pressure chamber 22 protective
membrane 1-8 thermocouples
Cross-section of the loop channel core center

• Fig. 19. Scheme of the capsule loop with the
  MGD-pump

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ILCC cross-section in the core central plane
Fig. 20. Scheme of the lead loop
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Main directions of investigation

• - Study of safety issues. A series of experiments
  on substantiation of fast sodium reactor safety
  was performed. Among them are feeding of gas
  into the core, sodium boiling, blocking of
  coolant flow in the experimental FA resulting in
  fuel elements damage, intercircuit leaks in steam
  generators etc. Detailed study of different
  normal and off-normal processes at the BOR-60
  reactor allowed testing and adjusting of methods
  and means of abnormities diagnostics.
• - Testing of fuel, absorbing and structural
  materials. Irradiation programs are paid special
  attention to, among them
• Mass testing of fuel elements and fuel assemblies
  up to the burn up of 30 h.a. under steady-state
  and transition conditions
• Testing of different neutron absorbing materials
• Radiation testing of structural reactor
  materials
• Testing of electric insulating, magnetic and
  refractory materials for fussion reactors

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• Investigations in radiation material science
• Determination of deformation, long-term strength
  and fracture toughness dependence at temperature
  of 320-1000?? up to the dose of 200 dpa
• Study of the technology of long-lived
  radionuclides transmutation and burning out from
  spent fuel of different reactors
• Radiation silicon alloying for radio electronics.
• In 1981 fuel elements with vibropacked fuel
  columns on the basis of power-generated plutonium
  were applied for the reactor core for the first
  time. Positive results of mass testing of fuel
  elements with vibropacked uranium-plutonium oxide
  fuel in the BOR-60 reactor up to the burn up of
  more than 30, as well as of 6 experimental fuel
  assemblies up to the burn up of 9,6 in the
  BN-600 reactor can serve a real basis for
  large-scale experiments in fast power reactors to
  increase their efficiency and to enhance their
  safety.

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• Testing of fuel elements containing weapon grade
  plutonium-based fuel was started in 1998. 
• In the frame of the program on development of
  closed fuel cycle elements much is being done on
  burning out and transmutation of plutonium and
  minor actinides (MA). Design-experiment
  investigations and analysis of the isotope
  content of microcapsules (40 pieces) with
  different MA sets irradiated in the BOR-60
  reactor were performed. The obtained
  design-experiment results can be used for
  adjustment of physical constants.

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• Results on investigation of different fuel
  compositions serve the basis for development of a
  fuel cycle of advanced fast reactors with
  enhanced safety. Among these is the BREST-OD-300
  reactor with lead coolant and nitride fuel.
• The first stage of testing of BREST-OD-300 pilot
  fuel elements took place at the BOR-60 reactor.
• Short-cut testing of different structural
  materials is carried out
• Steels used for fabrication of vessel internals
  (VI) for VVER reactors
• Zirconium alloys for VVER cores
• Vanadium-based alloys in lithium medium for
  fusion reactors
• Graphite for RBMK reactors.

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Table 6Reactor materials tested in the BOR-60
reactor
Material Material Type
Fuel Ceramics UO2, UO2-PuO2, UC, UN, UPuN, UPuCN
Fuel Metal U, UPu, UpuZrNb
Fuel Ceramal U-PuO2, UO2-U, UN-U
Absorbing Samples Ta, Hf, Dy, Sm, Gd, AlB6, AlB12, EuO3
Absorbing CPS rods CrB2, B4C, Eu2O3, Eu2O3H2Zr
Structural Stainless steels OX18H9, X18H10T, ??-450, ??-823 03?16?9?2, ??-912, ??-847, ??-172, ??-68, ??-24
Structural High-nickel alloys ??-16, ?20?45?4?, ???
Structural Refractory materials V, W, Mo, Nb
Structural Zirconium alloys ?-110, ?-635, ?-125
Structural Graphites ???-2-125, ??6-6, ??-280, ???, IG-11, ???
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Material Material Type
Electrotechnical Insulation Al2?3, SiO2, Si, mica
Electrotechnical Cables ????, ????(?)
Electrotechnical Magnets ????
Others Special ceramics ??-7, ??-46, ???, LiNbO3
Others Biological shielding materials Concretes
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Isotope accumulation for medical purposes

• Taking into account physical peculiarities of a
  fast reactor, commercial radionuclide
  accumulation parameters were investigated. The
  radionuclides were produced by the threshold
  neutron reactions 32P, 33P, 35S, 89Sr (reaction
  (n, ?)) and 117mSn (reaction (n,n')). Besides,
  indices of the 153Gd radionuclide accumulation
  process were also determined. The radionuclide
  was produced by reaction of radiation neutron
  capture (n,?) in the BOR-60 irradiation cells
  with specially heated neutron spectrum. At
  present serial production of strontium-89 from
  yttrium targets (for production of strontium-89
  without carrier preparation) and gadolinium-153
  from europium targets is realized for production
  of sources and preparations.

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Plans for future reactor facility operation

• The BOR-60 reactor has been in operation for 35
  years already, the design service life makes up
  20 years and calculated life is equal to 40
  years. Decision on possible reactor service life
  extension was made up taking into account the
  equipment and materials state, strength of the
  equipment and sodium circuit pipelines these
  are the components that contribute much to the
  reactor safety and that were fabricated in
  accordance with the current calculation norms.
  Long-term plans concerning the above mentioned
  problems are made for several decades.
• There are plans on reactor reconstruction aiming
  at the reactor service life extension for not
  less than 30 years in comparison with the
  calculated resource. During reconstruction it is
  important to expand the reactor experimental
  possibilities and to enhance its safety. A draft
  design of a new reactor has been prepared already
  and at present design work on installation of a
  new reactor within the operating reactor facility
  is being in process.

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Thank you for attention!