THE EXPECTATIONS OF THE FUEL VENDOR IN THE FIELD OF REACTOR MATERIAL SCIENCE OF NUCLEAR FUEL. THE RESULTS OF THE COOPERATION AND MATERIAL SCIENCE ASPECTS OF FUTURE RESEARCH IN THE MIDTERM PERSPECTIVE A.V. Ugryumov The 11th Conference on Reactor Material Science dedicated to the 55th Anniversary of the RIAR’s Reactor Material Testing Complex Dimitrovgrad, May 27-31, 2019

The main objectives of the development of nuclear fuel, taking into account the "Strategy-2018" of Rosatom The development of domestic nuclear generation and nuclear exports by increasing the efficiency and safety of VVER technology and introducing energy technologies based on fast neutron reactors The compliance with the increasing Customer requirements for safety and operational reliability of fuel and cores The improvement of the economic efficiency of various fuel cycles while ensuring an adequate level of reliability and safety of nuclear power plants in general Вестингауз бьется за рынок ВВЭР-440 в Европе. Два направления по заказу Операторов. В 2017 году нужно проработать и определить эффект.

Actual tasks for TVEL JSC as Russian Vendor of nuclear fuel Increased operational reliability Extended use of RT fuel Introduction of ATF Development of dry storage technology Justification of unit operation in load follow modes Increase of FA “energy storage capacity” Monitoring the operation of nuclear fuel at NPPs Improvement of structural and fuel materials Вестингауз бьется за рынок ВВЭР-440 в Европе. Два направления по заказу Операторов. В 2017 году нужно проработать и определить эффект.

R&D to improve and develop new nuclear fuels VVER and PWR Technology Improvement of nuclear fuel based on TVSA and TVS-2M for NPPs with VVER- 1000 in support of operation in 18-month cycles at a power level up to 107% Nном Justification and implementation of TVSA-T.mod.2 with increased uranium loading Implementation of TVSA-12 at Kozloduy NPP Development of the design of the universal fuel assembly TVS-2MT for VVER- 1000 and VVER-1200 units of Tianwan NPP Justification and implementation of the second generation antidebris filter (ADF-2) Justification of the safety of the transition of AES-2006 FA from the 4-year fuel cycle to the 18-month fuel cycle Improvement of nuclear fuel for VVER-1200 and VVER-TOI including justification of FAs with fuel rods without attachment in the base plate Improvement of FA designs using mixing grids, which allow to increase DNBR and increase power Justification of long-term dry storage of spent nuclear fuel Вестингауз бьется за рынок ВВЭР-440 в Европе. Два направления по заказу Операторов. В 2017 году нужно проработать и определить эффект.

R&D to improve and develop new nuclear fuels VVER and PWR Technology Justification of pilot operation of TVS-2M with REMIX-fuel Development of a stationary 18-month fuel cycle based on TVS-4 with fuel from reprocessed uranium taking into account isotopic composition of raw materials Development and implementation of TVS-K for NPP with PWR reactors Introduction of third-generation RK-3 fuel at Kola NPP Development of substantiating materials for the introduction of working assemblies RK-3+ without shroud at Dukovany NPP Reactor tests and pilot operation of three combined TVS-2M with ATF-type fuel rods at one of the units of Rostov NPP Reactor tests for justification of the operability of fuel rods of modern design in the projects of TVSA-12 for Kozloduy NPP, TVS-K, TVSA-Т.mod.2 for Temelin NPP Вестингауз бьется за рынок ВВЭР-440 в Европе. Два направления по заказу Операторов. В 2017 году нужно проработать и определить эффект.

R&D to improve and develop new nuclear fuels Fast Neutron Reactor Technology Development of BN-600 fuel assembly with fuel rod claddings of EK164 steel Development of CPS AR for BN-600 with end plugs of EK164 steel as part of the work on the transfer of BN-600 reactor core to the campaign of 592 EFPD Development of fuel rods and FA for BN-600 reactor core with the campaign of 800 EFPD Testing of prototype fuel rods for BN-1200 and BREST-OD-300 in experimental FA with full load of UPuN fuel Accident Tolerant Fuel R&D on production technology of sealed fuel rods of a new generation based on SiC Tests of experimental FA in MIR reactor Вестингауз бьется за рынок ВВЭР-440 в Европе. Два направления по заказу Операторов. В 2017 году нужно проработать и определить эффект.

R&D to improve and develop new nuclear fuels Improvement of Structural and Fuel Materials Post-irradiation examination of TVS-2M with fuel rod claddings made of new zirconium alloys E110M, E125 and E635 after operation during 3 fuel cycles at Unit 2 of Balakovo NPP Improvement of chromium-nickel alloy claddings (42XNM-type) in order to reduce their high-temperature embrittlement Reactor tests of new bimetallic claddings for FA of fast neutron reactors Preparation for testing of DUO-steel and vanadium alloys in BOR-60 reactor Reactor testing of experimental FA to achieve a damage dose of 120 dpa for EP823 and DUO-steel Ramp tests of TVS-K fuel rods with claddings from E110M, E110h.p. Post-irradiation examination and reactor tests of structural and fuel materials in research centers IFE Halden, Studsvik, ÚJV Řež, a.s. Вестингауз бьется за рынок ВВЭР-440 в Европе. Два направления по заказу Операторов. В 2017 году нужно проработать и определить эффект.

Cooperation with nuclear research centers Validation of the processes of design and licensing of Russian fuel in the national regulatory authorities Research on the properties and characteristics of structural and fuel materials Study and justification of the safety behavior of fuel rods of fuel assemblies of new designs under the conditions of RIA and LOCA Study of the operational behavior of nuclear fuel in standard FA designs Post-irradiation examination of spent nuclear fuel, reactor tests of new fuel Reactor tests of structural and fuel materials Halden Студсвик Řež Research and tests of structural and fuel materials Post-irradiation examination of structural materials Вестингауз бьется за рынок ВВЭР-440 в Европе. Два направления по заказу Операторов. В 2017 году нужно проработать и определить эффект.

Cooperation with JSC SSC RIAR Test and research programs with the participation of RIAR Program of post-irradiation examination of VVER fuel for 2016-2021 and up to 2025 Program of reactor tests of modern and advanced VVER and TVS-K fuel for 2013-2018 Reactor tests and post-irradiation examination in support of the development of FA with fuel rods of ATF-type for VVER and PWR Research program to justify the long-term dry storage of FA of new types

Post-irradiation examination of SFA in RIAR Program of post-irradiation examination of VVER fuel for 2016-2021 and up to 2025 Eight VVER-1000 FAs were examinated - six assemblies of the TVSA type and two of the TVS-2M type after operation at Kalinin, Balakovo and Rostov NPP with burnups from 24.4 to 57.6 MW·day/kgU Three TVSA had fuel rods with a thin cladding and a pellet of 7.8х0.0 mm One of FAs under study had fuel rods with claddings of E110h.p. sponge based alloy, and one of electrolytic based alloy. One of FAs used mixing spacer grids Seven FAs were identified at NPPs as leaking The obtained experimental data confirmed the design characteristics of the VVER-1000 FA and the new materials used The results were used to justify operation of modern design fuel rods with a pellet of 7.8х0.0 mm and U-Gd fuel rods with Gd2O3 content up to 8wt% and to perform licensing procedures with national regulatory authorities The results of studies of two CPS ARs after 10 years of operation at Unit 1 of Kalinin NPP allowed to justify an increase in the assigned life of ARs of this design up to 15 years The main case of FA leakage is defined as debris damage due to interaction with foreign objects The need to confirm the technical characteristics of VVER-1000 FAs with: Fuel rod claddings made of experimental alloys E110M, E125, E635, Burnup of 59.45 MW·day/kgU, Mixing grids like “Vihr” and “Progonka”, Unified top nozzle (TVSA-12PLUS), Fuel rods with pellets of 7.8х0.0 mm, 12 SGs and ADF, as well as determining the cause of leakage and justification of long-term storage of SFA in the storage pool required the updating of the current Program

Reactor tests of advanced fuel Program of reactor tests of modern and advanced VVER and TVS-K fuel for 2013-2018 According to the Program the following tests have been performed: Experiments with ramping power (RAMP): 16 full-scale fuel rods of TVSA VVER-1000 of various designs (with and without central hole, with different cladding thickness, with different grain size of fuel pellets) in the range of burnups from 43 to 75 MW·day/kgU 3 full-scale fuel rods of TVSA VVER-1000 with integrated gadolinium absorber with a burnup of 60 MW·day/kgU 3 experimental fuel rods of TVS-K design Experiments simulating RIA conditions: 12 refabricated fuel rods of various designs with claddings made of E110 and E110h.p. alloys and burnups from 30 to 75 MW·day/kgU Experiments simulating LOCA conditions: 3 refabricated fuel rods of various designs with claddings made of E110 and E110h.p. alloys and burnups from 43.9 to 76 MW·day/kgU Life tests in MIR reactor: 12 experimental fuel rods of TVS-K design with various cladding materials (E110M, E635M, E110h.p. and E125 alloys) 18 experimental fuel rods of FA AES-2006 design with various cladding materials (E110M, E635M, E110h.p., E125 and E110 alloys)

Reactor tests of advanced fuel (2) Thematically, the reactor experiments were aimed at: Determination of the life characteristics of fuel rods, including corrosion (FA AES-2006, FA PWR-2) Justification of the operability of fuel rods and U-Gd fuel rods in anticipated operational occurrences (AOO) Justification of fuel rod safety in design basis accidents Determination of the limits of the following design criteria: 1. Strength criterion ТС2 determining the critical value of gas pressure under the cladding 2. Strength criterion SC4 determining the limiting value of the relative strain of the fuel rod 3. Safety criterion for reactive accidents determining the limiting enthalpy of the fuel The main results of the reactor tests carried out under the Program were used to substantiate the operability of VVER / PWR fuel for foreign customers : - Fuel rod of FA VVER-1200 for Hanhikivi NPP (Finland) - Fuel rod of FA VVER-1200 for Paks-2 NPP (Hungary) - Fuel rod of TVSA-12 VVER-1000 for Kozloduy NPP (Bulgaria) - Fuel rod of TVSA-Т.mod.2 VVER-1000 for Temelin NPP (Czech Republic) - Fuel rod of TVS-K The obtained data allowed us to verify the RAPTA-5.2 code in terms of substantiating the thermomechanical behavior of fuel rods in the conditions of reactive accidents, as well as the kinetics of deformation of VVER and PWR fuel rods in LOCA conditions. The results were used in the certification of RAPTA-5.2 code for the burnup of 75.5 MW·day/kgU The obtained data allowed to verify the START-3A design code in the part of substantiation of the kinetics of fuel rod deformation, PCI, stress corrosion cracking of claddings, as well as fission gas release in transient modes and AOO modes of VVER and PWR fuel rods

Reactor tests of advanced fuel (3) The program has been implemented almost fully, given the focus of its implementation on the introduction of fuel with a thinned cladding and a fuel pellet without a central hole (TVSA-12 for Kozloduy NPP, TVSA-T.mod.2 for Temelin NPP, TVS-K) Actual tasks for which it is necessary to conduct research and experiments in the framework of the new program of reactor tests are as follows: 1. Justification of the compliance of the VVER and TVS-K fuel to the requirements of EUR: - conducting tests of fuel rods in load follow mode - conducting dynamic tests of irradiated fuel rods (substantiation of transport and technological operations in relation to dry storage of fuel) 2. Verification of the START-3A and RAPTA-5.2 codes: - study of the kinetics of fuel rod deformation in transient modes and in the AOO modes - study of the behavior of VVER-440 fuel rods with a burnup of 68 MW·day/kgU in LOCA conditions (the experiment has to be conducted as part of the Halden Reactor Project) 3. For the development of physical models of the behavior of fuel rods, it is necessary to conduct special experiments on the kinetics of fission gas release and fission gas swelling of fuel 4. To ensure fuel licensing of TVS-K in the regulatory authorities of Europe and the United States, reactor tests are required to determine the maximum allowable power before fuel melts 5. Studies of the behavior of VVER-100 fuel rods under controlled water chemistry conditions under operating conditions at power levels of 104% Nном and 107% Nном

Accident Tolerant Fuel Test in MIR reactor and post-irradiation examination Reactor tests and post-irradiation examination in support of the development of FA with fuel rods of ATF-type for VVER and PWR Two experimental FAs with samples of ATF fuel rods with VVER and PWR geometry were manufactured in 2018 and in January 2019 loaded for testing in experimental loops of MIR reactor with the corresponding water chemistry mode in order to provide experimental substantiation of the ATF technical design for VVER and PWR reactors Reactor tests of experimental FAs will be performed up to the burnup of ~ 50 MW·day/kgU The program provides for annual intermediate studies (visual inspection and profilometry) of experimental ATF fuel rods and the unloading of several fuel rods for post-irradiation examination The design of the MIR reactor allows simultaneous studies to be carried out in separate experimental loops, which is important considering the simultaneous testing of fuel for reactors of Russian and foreign design The first phase of reactor tests and post-irradiation examination of ATF-type fuel rods will be completed in 2019 The data obtained will be used to determine the optimal combination of fuel and cladding materials The next stage is the loading of experimental FAs with separate ATF fuel rods into the power reactor of a Russian NPP

Dry storage of SFA The purpose of the experiments in RIAR: Determination of changes in the length and diameter of the VVER-1000 fuel rod claddings with various designs and fuel burnup under conditions simulating technological operations for transferring from wet storage to dry storage, long-term normal dry storage and design basis accidents Pre-test studies of fuel rods. Testing of fuel rods in the vacuum drying mode and in the vacuum drying mode and design basis accidents Intermediate studies of fuel rods Tests of fuel rods in long-term dry storage mode (t=3500C, T=400-500 days) Nondestructive post-test studies of fuel rods Visual inspection Outer diameter measurements Eddy current testing Fuel rod length measurements Measurement of the thickness of the oxide film on the outer surface of cladding by the eddy current method Irradiated cladding samples are transferred to the NRC Kurchatov Institute for experimental studies to substantiate the dry storage of FAs of new types

Dry storage of SFA Using the results of experiments NRC Kurchatov Institute A set of experimental studies to substantiate the dry storage of FAs of new types include three parts: Tests to determine the creep rate and mechanical characteristics of the materials of fuel rod cladding in the initial and irradiated state The complex of microstructural studies in the initial, irradiated state and in the state after testing to determine the creep rate Tests to determine the threshold levels of hydride reorientation in the initial and irradiated state SSC RF TRINITI Development of a verified calculation code for predicting the behavior of fuel rods in dry storage conditions: Calculation of the fuel rod characteristics at the time of its unloading from the reactor to substantiate the dry storage Development of an improved assessment code for the calculation of the stress-strain state of the fuel rods of SFA under normal conditions and in case of accidents

Planned work on the adaptation (development) and verification of design models and design criteria for wet and dry storage conditions of fuel rods as part of SFA START-3A models Design criteria Planned works Experiments for model’s verification and development of criteria Stress-strain state + adaptation of the model and criterion for storage conditions RIAR in the framework of [1] Gas pressure inside cladding Creep NRC KI in the framework of [2] Reorientation of hydrides - development of the model and criterion VNIINM Grain boundary cracking development of the criterion VNIINM and RIAR Prolonged corrosion Stress corrosion cracking adaptation of the criterion for storage conditions Literature sources [1] Action Plan for meeting the requirements of the item 2.2.3. НП-061-05 regarding the establishment of an acceptable storage period of VVER-1000/1200 SFAs in the NPP design, Rosenergoatom, 2019-2022 [2] Program of experimental and calculation studies to substantiate the dry storage of new types of FAs, TVEL, NRC KI and TRINITI, 2015-2023

Tests of Russian materials in Halden Reactor Project Test object Test target Duration Test results 1 a) fuel with increased grain size (~25 μm); b) standard VVER fuel; c) standard uranium-gadolinium fuel of the VVER-1000 reactor Initial enrichment by U235 up to 10wt% Comparison with PWR fuel 2006- 2014 Burnup ~ 25 MW·day/kgU. Post-irradiation examinations were performed 2 E110 and E635 alloys Comparative corrosion tests (Zry-4, Zry-2, ZIRLO, M4, M5, E110 and E635) 1999-2009 Burnup ~ 50 MW·day/kgU. There are no significant differences between similar alloys 3 E110opt., E110M, E125opt. and E635M alloys Comparative corrosion tests in PWR conditions 2010- 2016 Test completed with positive results 4 Sponge-based E110opt. alloy Lift off tests 2014-2017 Test completed, report issued 5 Post-irradiation of VVER fuel from 50 to 72 MW·day/kgU Subsequent testing of the fuel rod in the LOCA conditions + RAMP test or fission gas release 2015-2018 Burnup currently achieved is 68 MW·day/kgU

Fuel rod tests in Studsvik Justification of operability of the TVS-K cladding material for Ringhals NPP Research on corrosion resistance, strength characteristics and resistance to deformation under irradiation in the conditions of the PWR reactor (non-destructive and destructive examination) The results of the work will be used in the licensing of TVS-K abroad (USA, Sweden, Finland) Fuel rod claddings from the TVS-K for Ringhals NPP after 3 cycles of operation up to burnups from 39.6 to 43.3 MW·day/kgU Experimental fuel rod claddings of various alloys from the IFA-728 fuel assembly after irradiation in Halden reactor loop during 900 EFPD up to burnups from 41 to 43 MW·day/kgU The length of each cladding was measured Visual inspection was conducted The claddings were punctured to determine the amount and composition of fission gas under the cladding A gamma scanning of the claddings was performed with the determination of fission product distribution The thickness of oxide film on the cladding outer surface was measured The fuel/cladding gap was measured SEM study of the claddings was conducted: - The morphology of the oxide film was studied - The distribution of the hydride phase in the metal was obtained - The elemental composition of the metal and oxide film was determined - TEM studies of the prepared samples are conducted, as well as continuing SEM studies of the claddings

Cooperation with ÚJV Řež, a.s. Agreement on cooperation in the planning, preparation and conducting reactor and post-reactor studies in the CVR experimental facilities, aimed at studying the properties of fuel compositions and structural materials, justifying the performance characteristics and confirming the implementation of new types of structural materials developed by TVEL JSC Research areas : Corrosion testing of structural materials of samples from alloys based on zirconium, nickel, steel and other alloys under development Tests in various water-chemistry modes, simulating operating conditions in reactors of various types Studies of properties and justification of operational characteristics of advanced fuels and structural materials Studies of fatigue characteristics, creep and mechanical tests of irradiated and non-irradiated products made of Zr-Nb alloys, Inconel alloys, steels and other materials in complex stress state conditions (under internal pressure and axial tension) Studies of the behavior (corrosion, hydrogenation) of materials without exposure and under the influence of radiation Carrying out strength calculations using the finite element method, modeling of the multi-axial stress state Experiments on dry storage – microstructural studies, hydride reorientation slow hydrogenated cracking of non-irradiated samples Irradiation in water with supercritical parameters

Conclusions The need to confirm the technical characteristics of VVER fuel of new designs requires updating the Program of post-irradiation examination for the medium term until 2025 The tasks set within the framework of the Reactor Test Program have been fulfilled almost fully, taking into account the particular focus of its implementation on the introduction of VVER-1000 fuel with a thinned fuel cladding and a fuel pellet without a central hole (TVSA-12 for Kozloduy NPP, TVSA-T. mod.2 for Temelin NPP, TVS-K) To meet the modern requirements of fuel Customers, TVEL JSC plans to conduct additional research and reactor experiments as part of the new Reactor Test Program Due to the shutdown of the Halden reactor, it is necessary to envisage the completion of pending experiments and research in other nuclear research centers In support of TVS-K project for Ringhals NPP and other foreign Customers, TVEL JSC will continue the post-irradiation examination of TVS-K fuel rods for design justifications of the new fuel For unconditional provision of Customers' requirements for fuel supply to countries focused on the technology of dry storage of SFAs, TVEL JSC will continue verification and certification of relevant design codes using the material science base of research centers TVEL JSC will continue the implementation of reactor experiments, modeling and development of design codes for justification of operability of accident tolerant fuel

Thanks for attention!