Current state and plans on validation of the mixed nitride fuel for fast reactors

1. Current state and plans on validation of the mixed nitride fuel for fast reactors M.V. Skupov, L.M. Zabudko May 28-30, 2019 JSC RIAR ХI conference on reactor materials science

2. Reactor aspects of the choice of nitride fuelfor FR • The appeal of dense fuel due to a possibility to develop reactors with core reproduction coefficient more than 1 • Operation of a reactor during an refueling intervals is carried out at the small, almost not changing reactivity, ~ 1 βэф • High heat conductivity of nitrides (10 timesabove oxides) allows to lower the operating temperature of fuel, to reduce power reactivity effect • The small accumulated energy in fuel (due to low thermal capacity) gives advantages at violations of normal operation • Upon transition of the closed nuclear fuel cycle (CNFC) the reactor is operated in the self-sufficiency mode, without feed by external plutonium, weapon grade plutonium does not accumulate • The decision to work hard on nitride fuel for perspective FR is made on October 29, 2009 on section No. 4 scientific and technical council of Rosatom by results of the analysis of options of dense fuel

3. The Russian experience and technological aspects of development of nitride fuel for FR • Long-term experience of manufacturing and operation of UN fuel in the BR-10 reactor: • Two loads of the reactor with UN (660 fuel elements and 590 fuel elements), the maximum burnup about 9 % h.a., the maximum linear power – 45 kW/m, the maximum fuel temperature 1175 K, density of pellets is 85-94 % t.d. • A large number of fuel elements tested in the BOR-60 reactor: • 1 FA with UN fuel, the maximum burnup more than 8 % h.a., maximum fuel temperature 1775 K; • Several fuel elements with (U,Pu)N (content of plutonium of 20%), the maximum burnup is4 % h.a. and 8.95 % h.a., the maximum temperature of fuel, respectively, 2475 K, 1750 K, density of pellets is 85-86 % t.d. • 2 ЭТВС с топливом (U,Pu)N и свинцовым подслоем, максимальное выгорание в одной из ЭТВС - 5,5 % т.а. • 2 FA with (U,Pu) N and a lead sublayer, the maximum burnup - 5.5% h.a. • 4 fuel elements with (U,Pu)N (high Pu content) in the frame of the Russian-French experiment of BORA-BORA, the maximum burnup of fuel is 12 % h.a. • Possibility of production of MOX and MNIT-fuel on universal installations • Possibility of reprocessing of the irradiated nitride fuel on the existing plants when carrying out some modification of the equipment

4. Preliminary key requirementsfor FR fuel elements with nitride fuel Fuel element for reactor BREST-OD-300 • Main requirements of the first operational phase: • Maximum burnup – 6 % h.a. • Maximum damage dose – 89 dpa • Maximum cladding temperature – 670 ° C • Maximum linear power – 42 kW/m • Cladding material is ferritic-martensitic steel EP823 Fuel element for reactor BN-1200 • Main requirements of the first operational phase: • Maximum burnup – 7.6% h.a. • Maximum damage dose – 96 dpa • Maximum cladding temperature – 680 ° C • Maximum linear power – 47 kW/m • Cladding material is austenitic steel EK164 • The option of fuel element with the reproducing material layer is considered • Increase of fuel burnup at the subsequent stages is planned using ferritic-martensitic steel and its ODS modifications

5. Out-of-pile properties:Thermochemical stability of (U,Pu)N The evaporation process is noticeable at 1800 °C Before After 2000 оС T, C Ion current*10-11/A mg/min Temperature, mg Mass change, mg mg Time/min Maximum observed depletion of plutonium as a result of tests – up to 2 times Time/min

6. Program of validation of nitride fuel elements for starting cores

7. Test results of fuel rods with a maximum burnup of 7.5 % t. a. Post-irradiation examination of CEFA-7 (4 mcc; 6,9x0,4 mm CnS68; 7.5 % t.a.; 73.6 dpa) Pellet center Mid of fuel column Top of fuel column - The rate of fuel swelling in different sections of the fuel column is from 1.5 % to 2.1 % at 1 % h.a. - The maximum value of the relative increase in the diameter of fuel rods with MNIT fuel is 0.9 %, which is less than the same value for oxide fuel rods – 1.7 %

8. Post-irradiation examination of BN-1200 prototype fuel rods (CEFA-2) 4 mcc; EK164; 5,05 % h.а.; 62,1 dpa Elongation: 1 to 5,6 mm The relationship of elongation, the gas releaseand ovalization is revealed Diameter, mm Diameter, mm Distance from bottom of fuel rod, mm Distance from bottom of fuel rod, mm Distance from bottom of fuel rod, mm MOX fuel rod №30 MNIT fuel rod №22 MNIT fuel rod №31

9. Post-irradiation examination of BN-1200 prototype fuel rods (EFA-4) 4 mcc; EK164; 5,0 % h.а.; 58,5 dpa Segment sample Mid cross-section Typical corrosion of clads made of EK164 steel Tensile strength, MPa Mass fraction, % Elongation, % Temperature, oС Temperature, oС Carbone Nitrogen Ringsamples Segmentsamples Ringsamples Segmentsamples The resource of the fuel rods is not depleted, maximum corrosion damage of up to 60 µm

10. Post-irradiation examination of BREST prototype fuel rods (CEFA-3) 4 mcc; EP823; 4,54 % h.а.; 53,0 dpa The relationship of elongation, the gas releaseand ovalization is confirmed Elongation: 1 to 5,3mm Diameter, mm Diameter, mm Diameter, mm Distance from bottom of fuel rod, mm Distance from bottom of fuel rod, mm Distance from bottom of fuel rod, mm UO2fuel rod №30 MMNIT fuel rod №26 MNIT fuel rod №30

11. Post-irradiation examination of BREST prototype fuel rods (EFA-5) 4 mcc; EP823; 3,80 % h.а.; 47,5 dpa Sample from «cold» partof fuel rod Mid cross-section Depthof penetration of corrosion, mm * Oxygen content Typical picture of observed corrosion Oxygen content, %

12. Current state of (U,Pu)N validation on the following criteria: burnup and the damage dose BN-600 BOR-60 Plan for 2024 Material FA BN-1200 BREST-OD-300 Damage dose, dpa Burnup, h.a.

13. MNIT fuel burnup enhancement program The transition to a new test method, which provides for the operation of fuel rodsup to the limit parameters, is carried out IA- irradiation assembly IPS – in-pile storage PIE – post irradiation examination

14. A sketch of the irradiation assembly fortesting the fuel limits (OKBM) Irradiation assembly Removal conteiner Ampoule with fuel rod The use of the leak node worked on absorbing elements of BN reactors

15. Current availability of calculation codes for fuel rods with oxide and nitride • The results of project «Dense fuel»: code CORAT in preparation for recertification • An improved correlation code of fuel rod developer • It used for technical design and design studies of fuel rods with oxide and nitride fuel • The results of project «Engineering codes»: code DRACON at the stage of certification • Correlation code with the inclusion of mechanistic models • It used for confirming calculations and design studies of fuel rods with MNIT fuel in technical design • The results of project «New generation codes»: engineeringcodeBERCUT was certified • Correlation code with the inclusion of mechanistic models • It used for test calculations and design studies of fuel rods with MNIT fuel • The results of project «New generation codes»: mechanisticcodeBERCUT was developed • Mechanistic code with elements of correlation • At the stage of testing CORAT-designercode interface

16. Irradiation testing in research reactors • The program of ampoule tests of MNIT fuel in the IVV-2M reactor for verification of the BERCUT mechanical code models using: • Blown experimental device with simultaneous periodic monitoring of the composition and kinetics of release of the short-lived gaseous fission products to achieve the fuel burnup of ~ 1 %h.a. • Three irradiation devices with sealed vials to achieve the fuel burnup of ~1.5; ~2.5; ~3.5 % h.a. • Manufactured irradiation device for IVV-2M. Irradiation tests – 2019-2020 • In the reactor BOR-60 continue to test fuel rods with MNIT fuel: • ID-2 – prototypefuel elements BREST (gas sublayer) • ID-2 – prototypefuel elements BREST (lead sublayer) • ID-10 – fuel elements withEK181 и ChS139 claddings (BN-1200 steels) • ID-16 – fuel elements withbimetallic cladings on the basis of EP900 and fuel rods with MA • The test under emergency conditions in the reactors IGR and the MIR are preparing а) b) а) – leaky ampoule; b) – sealed ampoule

17. Conclusion • The pre-reactor properties of nitride fuel are obtained in a volume sufficient for technical design, however, further development of the hardware and methodological base is required to improve the accuracy characteristics of the calculations. The firstrelease of thecertification report on properties of the MNIT fuel - 2019. • Positive results of testing of prototype fuel rods BN-1200 and BREST-OD-300 in FA with full load of MNIT fuel were obtained. In RIAR tested and implemented the most adequate testing methods using tube samples, along with circular and segmental samples. The data will be used to extend the tests of the remaining FA BN-600 to validate the fuel rods for the first loads. • Special attention in the development of calculating codes will be given to the transitional modes and emergency modes. • Significant results for new types of claddings were obtained: fuel rods with bimetal claddings are already testing, ODS steels and vanadium alloys are being prepared for testing in BOR-60. Testing of the materials science assembly to achieve a damage dose of 145 dpa on EP823 and ODS steels. • Increasing the burnup of the MNIT fuel with new materials will be carried out in theBN-600 with the use of CEFA and new testingtools – irradiation assemblies BN-600.