A A Bochvar Hightechnology Research Institute of Inorganic

A. A. Bochvar High-technology Research Institute of Inorganic Materials (SC «VNIINM» ) «ROSATOM» STATE ATOMIC ENERGY CORPORATION E 110 M ALLOY FUEL ROD CLADDINGS IN-REACTOR TESTS IN WATER-COOLED REACTORS AND POST-IRRADIATION EXAMINATION RESULTS A. YU. SHEVYAKOV, V. A. MARKELOV, V. V. NOVIKOV, N. S. SABUROV, A. YU. GUSEV, V. F. KON’KOV, M. M. PEREGUD SC «VNIINM» , MOSCOW, RUSSIA ХI conference on reactor materials science dedicated to the 55 th anniversary of the RIAR reactor materials science department May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

Introduction The development and modernization of Zr alloys for fuel rod claddings continues to receive attention from leading fuel suppliers. The main goal of this work is to enhance corrosion and creep resistance of the zirconium claddings. The Halden Reactor (HR) LWR loops have been widely used in international practice for evaluating of corrosion characteristics of Zr alloys. Halden Reactor This paper presents the results of the bilateral project performed in the HR where the fuel test assembly IFA-728 with experimental fuel claddings from Russian alloys (E 110 opt, E 110 M, E 125 and E 635 M) has been tested under high Li PWR water chemistry regime (WCR) for direct comparison of the corrosion resistance, hydrogenation and irradiation creep. In addition, some results from irradiation of similar cladding samples from these alloys in the BOR-60 reactor on diametric creep under internal pressure also presented. ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 2

Test materials Claddings with outer diameter of 9. 5 mm and wall thickness of 0. 57 mm from E 110 opt, E 110 M, E 125 and E 635 M alloys, which compositions are presented in the table, have been used Сплав Nb, % Sn, % Fe, % O, % E 110 opt 1. 05 - 0. 055 0. 085 E 110 M 1. 02 - 0. 095 0. 120 E 125 2. 45 - 0. 035 0. 069 E 635 M 0. 79 0. 81 0. 335 0. 075 Experimental fuel rods with fuel column length of 200 mm were tested in the HR PWR loop 200 mm ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 3

Test conditions in Halden Reactor Upper cluster (UCl) Ø 5 – E 110 M Ø 6 – E 125 Ø 7 – E 635 М Ø 8 – E 110 opt Lower cluster (LCl) Ø 1 – E 110 M Ø 2 – E 125 Ø 3 – E 635 М Ø 4 – E 110 opt WCR characteristic: Test conditions: Ø Li: 9, 2 – 10, 6 ppm Ø Full power days: 907 eff. days Ø B: 1524 – 1702 ppm Ø Burn up: ~ 60 MW∙day/kg. U Ø H 2: 2 – 3, 5 ppm Ø Cladding temperature: 351°С Ø р. Н 300 – 7, 4 ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 4

Reactor and post-irradiation examinations methods During irradiation intermediate eddy-current oxide film thickness measurements in four sections of experimental fuel rod claddings placed in the upper cluster was performed. After finishing irradiation, the inspection and destructive tests were carried out including: ü visual assessment, photographing and eddy-current measurement of the oxide layer thickness in four sections of the equipped samples of claddings located in both clusters; ü diameter tracing measurement conducted along the fuel rods using three-pronged inductive sensor; ü elongation measurement by the distance between the pre-marked indicators on the fuel rod lower and upper plugs; ü metallographic analysis of structure and the oxide film thickness, distribution and orientation of hydrides; ü determination of hydrogen content by high-temperature extraction method. ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 5

Autoclave tests In addition, a set of the fresh samples made from the same types of the cladding materials was tested in the autoclave at similar to IFA-728 water chemistry and cladding temperatures: § 9 – 12 ppm Li § 1800 ppm B § 350 ºC § 18, 6 МPа Duration of the autoclave test was 930 days. The obtained results were compared to the Halden test IFA 728 to study effects of the irradiation on the corrosion and its acceleration. ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 6

Results on corrosion Eddy current measurements of the oxide film thickness of the upper cluster fuel claddings after each irradiation cycle Oxide film thickness, μm Reactor tests Full power days Burn up, MW∙day/kg. U E 110 M E 125 E 110 opt E 125 E 635 M Oxide film thickness, μm Autoclave tests E 635 M Time, days E 110 opt E 110 M ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR E 110 opt E 125 E 635 M 7

Results on corrosion Metallography measurement results of oxide films after autoclave and reactor tests Alloy Average oxide thickness (h), μm Reactor E 110 opt 10. 2 18. 3 E 110 M 11. 1 E 125 10. 6 Number of layers in oxide Average layer thickness in oxide, μm Autoclave Reactor 1. 8 5 6 2. 2 3. 0 18. 2 1. 6 5 6 2. 2 2. 9 13. 6 1. 3 4 4 2. 6 3. 2 E 635 M 14. 7 48. 8 3. 3 h. R/h. A – oxide thickness ratio in reactor and in autoclave; * – it was impossible to determine 7 -* 2. 0 -* Reactor tests Autoclave h. R/h. A ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 8

Results on hydrogenation Hydrides distribution Hydrogen content , ppm Determination of hydrogen content by high-temperature extraction method Autoclave tests Oxide film thickness, μm E 110 opt E 125 E 635 M Re ac tor te s ts Eltra OH 900 E 110 M ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 9

Dimensional measurements results Geometric dimensions measurement of the fuel rod claddings after irradiation: elongation and diameter change Diameter, mm E 635 M E 125 E 110 opt E 110 M Elongation, % Initial cladding diameter – 9. 5 mm High level, mm Elongation, % LCl E 110 M E 110 opt E 125 E 635 M v Radiation-thermal creep deformation under the influence of thermo-hydraulic conditions UCl v Irradiation growth deformation v Creep deformation after pellet-cladding interaction Burn up, MW∙day/kg. U E 110 M E 110 opt E 125 E 635 M ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 10

Results on creep under internal pressure Tangential deformation, % BOR-60 research reactor Fluence (BOR-60), × 1022 cm-2 (E ≥ 0, 1 Мe. V) E 110 M o Diameter stresses of creep of specimens under internal pressure: 100 МPа o Irradiation temperature: (315 ÷ 325) °C o Neutron fluence: 5, 4× 1022 cm-2 (E ≥ 0, 1 Мe. V) E 110 opt E 125 E 635 M ØThe dependence of the tangential deformation of gas-filled samples on the irradiation time is approximated by a linear law. ØCladding creep rate: • E 110 М ~ 1, 2× 10 -4 %/h • E 110 opt ~ 1, 5× 10 -4 %/h • E 125 ~ 1, 8× 10 -4 %/h • E 635 М ~ 0, 5× 10 -4 %/h ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 11

Conclusions Test results in Halden Reactor under advanced PWR conditions with high Li concentration (up to 10 ppm) and PIE have shown that: ü The best corrosion and hydrogenation resistance but the worst elongation creep resistance under irradiation was observed on E 125 fuel rod claddings. The worst corrosion and hydrogenation resistance with the best elongation resistance under irradiation was observed on E 635 M fuel rod claddings; ü The optimal combination of corrosion, hydrogenation and elongation resistance in reactor was observed in E 110 opt and E 110 M fuel rod claddings. At the same time, E 110 M has the better resistance to irradiation creep comparing with E 110 opt with practically similar corrosion. ХI conference on reactor materials science May 27 -31, 2019, Russia, Dimitrovgrad, SC SSC RIAR 12

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