E110 Alloy Cladding Tube Properties and Their Interrelation with Alloy Structure-Phase Condition and Impurity Content

Abstract

The E110 alloy (Zr-1Nb) designed in Russia has been in successful operation in the VVER and RBMK reactors since 1960. The evolution of the Russian reactors during this period of time, their numbers and operating parameters, and, hence, the relevant production of the cladding tubes required much effort from the researchers and producers to optimize the technological processes resulting in the high operating reliability of the fuel rods clad in the E110 alloy. In the Zr-Nb system to which the E110 alloy pertains, highly sophisticated structure-phase transformations proceed depending on the temperature-rate conditions of processing and the impurity content of the alloy. The alloy structure formed during its processing affects substantially the processability and performance of semi-products and finished tubes. The paper demonstrates the interrelation between the structure-phase transformations of the E110 alloy and the temperature-rate conditions of processing as well as the oxygen content of the alloy. The phase composition and microstructure are shown to play an essential part in providing the tubes with the requisite properties and their stability. The properties of cladding tubes having different structures and fuel rod claddings as in-pile irradiated are analyzed. The corrosion behavior of E110 alloy claddings is shown to be strongly dependent upon the material structure and reactor operating conditions.

High-temperature steam-oxidation behavior of Zr-1Nb cladding alloy E110

“…The hydrogen concentration for this as-received tubing was <10 wppm. Table 1 lists the measured concentrations of major alloying elements and some impurities in E110 used in the ANL study, along with the E110 specifications reported by Shebaldov et al [12]. Although Fe, Sn and C concentrations are higher than ''typical" values, while Hf is lower than the typical value, the chemistry of the Fortum-supplied E110 is essentially within E110 specifications.…”

“…As such, extensive studies have been conducted over the past decades on the hightemperature steam-oxidation of fuel-rod cladding made of traditional zirconium alloys, such as Zircaloy-2 (Zry-2) and Zircaloy-4 (Zry-4) [1][2][3][4][5][6][7][8][9]. Advanced zirconium alloys, such as Zr-Sn-Nb and Zr-Nb, exhibit better corrosion resistance than Zry-4 under the operating conditions of pressurized water reactors (PWRs) [10][11][12][13][14][15][16][17][18]. Cladding alloys such as ZIRLO TM (Zr-1 wt.% Sn-1 wt.% Nb) and M5 TM (Zr-1 wt.% Nb) have replaced Zry-4 in most US PWRs, and E110 (Zr-1 wt.% Nb) is used in PWRs designed by the Russian nuclear vender (TVEL).…”

Cladding embrittlement during postulated loss-of-coolant accidents.

“…d J-P. Mardon et al [6]. e P. V. Shebaldov et al [7]. f Oxygen is considered an alloying element in these alloys.…”

“…The dimensional and chemical characterization results are listed in Table 6 for E110 and Table 5 for M5. For the alloying and impurity elements measured, materials used in the ANL test program are within specifications for E110 [7] and M5 [6]. Noticeable differences in materials are: thicker wall for E110 (0.71 mm vs. 0.57-0.61 mm for M5), higher surface roughness for E110 tubing (0.34 µm vs. 0.12 µm for M5), and lower oxygen content for E110 (0.05 wt.% vs. 0.145 wt.% for M5).…”

Recent Advances in Protective Coatings for Accident Tolerant Zr-Based Fuel Claddings

“…However, zirconium alloys demonstrate poor oxidation kinetics at elevated temperatures. Since 2011, following the tragic Fukushima Daiichi Nuclear accident, great attention has been devoted to the development of a new concept of nuclear fuel to improve the safety of nuclear reactors during normal operation, transient modes and under accident conditions [5][6][7][8][9][10]. A concept called "accident tolerant fuel (ATF)" indicates a strategy to prevent/limit the interaction of cladding material with water steam, or hydrogen embrittlement, and to reduce heat generation during cladding oxidation and increase "processing time" under accident conditions before re-flooding of the nuclear core [9,11,12].…”