1. A.A. Bochvar High-technology Research Institute of Inorganic Materials (SC «VNIINM»)
«ROSATOM» State Atomic Energy Corporation
E110M 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 55th anniversary of the RIAR reactor materials science department
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

2. 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 (E110opt, E110M, E125 and E635M) 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

3. Test materials
Claddings with outer diameter of 9.5 mm and wall thickness of 0.57 mm from E110opt, E110M, E125 and E635M alloys, which compositions are presented in the table, have been used
Сплав
Nb, %
Sn, %
Fe, %
O, %
E110opt
1.05 -
0.055
0.085
E110M
1.02
0.095
0.120
E125
2.45
0.035
0.069
E635M
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

4. Test conditions in Halden Reactor
Upper cluster (UCl)
5 – E110M
6 – E125
7 – E635М
8 – E110opt
Lower cluster (LCl)
1 – E110M
2 – E125
3 – E635М
4 – E110opt
WCR characteristic:
Li: 9,2 – 10,6 ppm
B: 1524 – 1702 ppm
H2: 2 – 3,5 ppm
рН300 – 7,4
Test conditions:
Full power days: 907 eff. days
Burn up: ~ 60 MW∙day/kgU
Cladding temperature: 351°С
ХI conference on reactor materials science
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

5. 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:
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;
diameter tracing measurement conducted along the fuel rods using three-pronged inductive sensor;
elongation measurement by the distance between the pre-marked indicators on the fuel rod lower and upper plugs;
metallographic analysis of structure and the oxide film thickness, distribution and orientation of hydrides;
determination of hydrogen content by high-temperature extraction method.
ХI conference on reactor materials science
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

6. 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

7. Oxide film thickness, μm Oxide film thickness, μm
Results on corrosion
Eddy current measurements of the oxide film thickness of the upper cluster fuel claddings after each irradiation cycle
Reactor tests
Burn up, MW∙day/kgU
Full power days
Oxide film thickness, μm
E110M
E110opt
E125
E635M
E110M
E110opt
E125
E635M
Autoclave tests
E110M
E110opt
E125
E635M
Oxide film thickness, μm
Time, days
ХI conference on reactor materials science
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

8. Results on corrosion
Metallography measurement results of oxide films after autoclave and reactor tests
Alloy
Average oxide thickness (h), μm
hR/hA
Number of layers in oxide
Average layer thickness in oxide, μm
Autoclave
Reactor
E110opt
10.2
18.3
1.8 5 6
2.2
3.0
E110M
11.1
18.2
1.6
2.9
E125
10.6
13.6
1.3 4
2.6
3.2
E635M
14.7
48.8
3.3 7 -*
2.0
hR/hA – oxide thickness ratio in reactor and in autoclave;
* – it was impossible to determine
Autoclave tests
Reactor tests
ХI conference on reactor materials science
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

9. Results on hydrogenation
Determination of hydrogen content by high-temperature extraction method
Hydrides distribution
Oxide film thickness, μm
Hydrogen content , ppm
E110M
E110opt
E125
E635M
Autoclave tests
Reactor tests
Eltra OH900
ХI conference on reactor materials science
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

10. Dimensional measurements results
Geometric dimensions measurement of the fuel rod claddings after irradiation:
elongation and diameter change
Elongation, %
E110M
E110opt
E125
E635M
E110M
E110opt
E125
E635M
Diameter, mm
High level, mm
Initial cladding diameter – 9.5 mm
UCl
LCl
Burn up, MW∙day/kgU
E110M
E110opt
E125
E635M
Elongation, %
Radiation-thermal creep deformation under the influence of thermo-hydraulic conditions
Irradiation growth deformation
Creep deformation after pellet-cladding interaction
ХI conference on reactor materials science
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

11. Results on creep under internal pressure
BOR-60 research reactor
E110opt
E125
E635M
E110M
Fluence (BOR-60), ×1022 cm-2 (E ≥ 0,1 МeV)
Tangential deformation, %
Diameter stresses of creep of specimens under internal pressure: 100 МPа
Irradiation temperature: (315 ÷ 325) °C
Neutron fluence: 5,4×1022 cm-2 (E ≥ 0,1 МeV)
The dependence of the tangential deformation of gas-filled samples on the irradiation time is approximated by a linear law.
Cladding creep rate:
E110М ~ 1,2×10-4 %/h
E110opt ~ 1,5×10-4 %/h
E125 ~ 1,8×10-4 %/h
E635М ~ 0,5×10-4 %/h
ХI conference on reactor materials science
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

12. Conclusions
Test results in Halden Reactor under advanced PWR conditions with high Li concentration (up to 10 ppm) and PIE have shown that:
The best corrosion and hydrogenation resistance but the worst elongation creep resistance under irradiation was observed on E125 fuel rod claddings. The worst corrosion and hydrogenation resistance with the best elongation resistance under irradiation was observed on E635M fuel rod claddings;
The optimal combination of corrosion, hydrogenation and elongation resistance in reactor was observed in E110opt and E110M fuel rod claddings. At the same time, E110M has the better resistance to irradiation creep comparing with E110opt with practically similar corrosion.
ХI conference on reactor materials science
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR

13. THANK YOU FOR YOUR ATTENTION!
ХI conference on reactor materials science
May 27-31, 2019, Russia, Dimitrovgrad, SC SSC RIAR