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

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.

Table 1 Some physical characteristics of the reactor Table 1 Some physical characteristics of the reactor CharacteristicValue Reactor heat power, MW60 Inlet temperature of coolant,  С Outlet temperature of coolant,  С 530 FuelUO2 or UO2-PuO2 235U enrichment, %45-90 Maximum Pu concentration, %40 Maximum volumetric power in the core, kW/l1100 Maximum neutron flux density, cm -2 ·s ·10 15 Average neutron energy, MeV0.4 Neutron fluence per 1 year, cm -2 3·10 22 Damage dose accumulation rate, dpa/yUp to 25 Fuel burnup rate, %/yUp to 6 Power non-uniformity factors: Axial Radial

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

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 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

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

Fig. 4. Cartogram of the BOR-60 reactor Reactor loading possibility Cells quantity for S/A for absorbing rods instrumented cells State S/A quantity 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

Fig. 5. Radial distribution of average neutron energy (En), integral energy (Fn) and neutron flux density with Е>0.1 Mev (Fn(0.1))

Fig. 6. Neutron spectrum of the BOR-60 reactor core - layer (cell number)

Fig. 7. Neutron spectrum of the BOR-60 reactor reflector – layer (cell number)

- 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 (А43 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.

Table 2 Testing conditions of materials and products in cell D-23 ParameterValue Neutron flux density, sm -2 ·s -1 2·10 15 Specific radiation energy release in structural materials (with atomic number Z = 26  30), W/g 4 Absorbed gamma-radiation dose rate, Gy/s4.5·10 3 Coefficient of non-uniform radiation density distribution along the core height (450 mm): for neutrons for gamma-radiation Sodium flow rate, m 3 /h: when fed from high pressure chamber when fed from low pressure chamber up to 8 up to 2

Table 3 Neutron-physical characteristics of the BOR-60 instrumented cells (Wreactor=55 MW) Table 3 Neutron-physical characteristics of the BOR-60 instrumented cells (Wreactor=55 MW) 1 year of irradiation - WT≈ MW×h, Kz and Kr – axial and radial non-uniformity coefficient. Cell, rowЕ31, 1А43, 3D23, 5D35, 8 Radius of the cell center location against the core center, mm Neutron flux density, sm -2 s -1 : -E>0.0 MeV (F0) -E>0.1 MeV (F0.1) Damage accumulation rate in steel (DPA), d.p.a./s K z (AP), relative unit F F DPA K r (CCP), relative unit F0F DPA Neutron flux density fraction with energy exceeding 0.1 MeV, relative unit Average neutron energy, keV Neutron fluence, sm -2 E>0.0 MeV E>0.1 MeV Steel damage dose, d.p.a

Fig. 8. Neutron spectrum of cell G01 of the BOR-60 reflector

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 1 - HEC, 2 - sand, 3 - oxide, 4 – disperser drive, 5 – cast iron, 6 - graphite, 7 - concrete, 8 - VEC

Fig. 10. Neutron spectrum of the BOR-60 VEC (calculations were made on the basis of MMK and OKS-ROZ-6 programs)

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 -2 s -1 ) Table 4 Neutron flux and gamma-quantum density at the BOR-60 HEC outlet (sm -2 s -1 ) HEC Calculated value Experiment value Fn FF Without Pb-screen, Еn>0 MeV (0.84  1.2)   10 8 (2.9  3.4)  10 8 Without Pb-screen, Еn>1.2 MeV (6.2  8.6)  (5.7  6.5)  10 7 With Pb-screen 3.6  

Fig. 12. Typical diagram of the BOR-60 reactor operation

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.

Table 5 Irradiation parameters errors, % ParameterMeasurementCalculation Intel (outlet) reactor temperature 1,2- Reactor power-2,5 Reactor flow rate3- Neutron flux (fluence)710 Experimental devices (ED) flow rate 2- ED power710 Intel ED temperature11,5 Outlet ED temperature11-3

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.

Fig. 13. Flow experimental assemblies Fig. 14. Device with evaporative with gas heat insulation: thermosiphon: 1,2 – outer and inner bodies; 3 – clearance; 4 – shells; 5 - specimens 1,2 – outer and inner bodies; 3 – clearance; 4 – shells; 5 - specimens 1 – specimens; 2 – shell; 3 – heater; 4 – body 1 – specimens; 2 – shell; 3 – heater; 4 – body

Fig. 15. Dismountable assembly with a hot probe for irradiation of structural materials 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 The lower boundary of the irradiation temperature range that is ensured in the BOR-60 reactor makes up о С. 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.

Fig.16. Cross-section of the experimental device with capsules for irradiation of vanadium in lithium medium 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 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. 17. Scheme of sodium boiling generator 1 - fixture 2 - throttling orifice 3 - filter 4 - block of tungsten rods 5 - gas clearance 6 - block of steel rods 7 - head 1 - fixture 2 - throttling orifice 3 - filter 4 - block of tungsten rods 5 - gas clearance 6 - block of steel rods 7 - head

Fig. 18. Scheme of the instrumented nozzle 1 – fuel assembly 2 - nozzle body 3 – tube with sensors 4 – flow regulator 5 – sodium vapor filter 6 – electric engine 1 – fuel assembly 2 - nozzle body 3 – tube with sensors 4 – flow regulator 5 – sodium vapor filter 6 – electric engine

Fig. 19. Scheme of the capsule loop with the MGD-pump 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 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. 20. Scheme of the lead loop ILCC cross-section in the core central plane

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;

Investigations in radiation material science: Determination of deformation, long-term strength and fracture toughness dependence at temperature of о С 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.

Testing of fuel elements containing weapon grade plutonium-based fuel was started in 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.

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.

Table 6 Reactor materials tested in the BOR-60 reactor MaterialType Fuel Ceramics UO 2, UO 2 -PuO 2, UC, UN, UPuN, UPuCN MetalU, UPu, UpuZrNb CeramalU-PuO 2, UO 2- U, UN-U Absorbing Samples Ta, Hf, Dy, Sm, Gd, AlB 6, AlB 12, EuO 3 CPS rodsCrB 2, B 4 C, Eu 2 O 3, Eu 2 O 3 +H 2 Zr Structural Stainless steels OX18H9, X18H10T, ЭП-450, ЭП Х16Н9М2, ЭП-912, ЭИ- 847, ЭП-172, ЧС-68, ВХ-24 High-nickel alloysРЕ-16, Х20Н45М4Б, ВЦУ Refractory materials V, W, M o, N b Zirconium alloysЭ-110, Э-635, Э-125 Graphites ГРП-2-125, МП6-6, ГР-280, АРВ, IG-11, ПГИ

MaterialType ElectrotechnicalInsulationAl 2 О 3, SiO 2, Si, mica CablesКТМС, КНМС(Н) MagnetsЮНДК OthersSpecial ceramicsГБ-7, ИФ-46, ЦТС, LiNbO 3 Biological shielding materials Concretes

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: 32 P, 33 P, 35 S, 89 Sr (reaction (n, р)) and 117m Sn (reaction (n,n')). Besides, indices of the 153 Gd 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.

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. Plans for future reactor facility operation

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