Mechanical Properties of Zr-2.5Nb Pressure Tubes Made from Electrolytic Powder

Abstract

Pressure tubes in CANDU and RBMK reactors are fabricated from Zr-2.5Nb alloy. This paper describes the mechanical properties of tubes used in power reactors made by four routes using electrolytic powder as the base material. The microstructures developed by each route are distinct: CW: cold-worked material consists of flattened α-Zr grains surrounded by a skin of β-Zr phase; used in CANDU 6 reactors. CW-A: material that was cold-worked and annealed at 540°C contains elongated α-Zr grains mixed with equiaxed α-Zr grains and particles of β-Nb phase; used in all RBMK 1000 reactors. TMT-1: material quenched from the (α+β)-Zr phase into water follow by cold-working consists of α'-phase and between 10 and 20 % of untransformed α-phase; used in RBMK 1500, Ignalina 1. TMT-2: material quenched from the (α + β)-phase into argon-helium gas mixture followed by cold-working consists of Widmanstätten α-phase and untransformed α-phase. This material is used in RBMK 1500, Ignalina 2. The CW and TMT-2 tubes have a higher proportion of grains with basal plane normals in the transverse direction, FT of 0.52 to 0.57, than in the radial direction, FR of 0.38, while quenched and annealed materials (TMT-1 and CW-A) have similar values of FT and FR, about 0.38 in quenched materials and 0.41 in annealed materials. Transverse tensile strength, crack growth resistance, dJ/da, and axial crack velocity, VH, of delayed hydride cracking (DHC) were evaluated, using standard techniques, between 250 and 300°C on as-fabricated materials. In-reactor creep deformation was evaluated from measurements of tube diameter in RBMK 1000s, RBMK 1500s and two CANDU 6 power reactors. Strength and crack growth resistance were measured on TMT-1 and TMT-2 tubes removed from Ignalina NPP Units 1 and 2 after 12–17 years of in-reactor service. As-received cold-worked material had the highest strength; the annealed material had the lowest strength while the quenched materials had intermediate strength. Irradiation increased the strength by about 200 MPa in all four materials. Although DHC is sensitive to texture and the distribution of the β-Zr phase, the dominating factor controlling crack velocity appears to be material strength: with an increase of strength by a factor of about two, VH increased by a factor of 30. Since harmful trace elements were well controlled during manufacturing, other factors affecting crack growth resistance could be assessed. Again, strength appeared important; dJ/da declined approximately linearly with increase in strength induced by irradiation, decreasing from about 350 to 100 MPa as the strength increased from about 250 to 850 MPa. The exception was TMT-2 material where crack growth resistance was maintained after irradiation. TMT-2 material also had good diametral creep resistance in-reactor, attributed to both its texture and grain structure. The other three materials had similar creep resistance controlled mostly by their texture.

Material development for India’s nuclear power programme

“…The cold worked and annealed tubes and gas quenched and aged tubes have much lower strength and do not meet the short term strength requirement of the pressure tubes for Indian heavy water reactors. Cold worked and stress relieved Zr-2.5 Nb pressure tubes reportedly show low inreactor dimensional changes at lower fluences and shows accelerated behaviour at larger fluences (Coleman et al 2007). On the other hand, heat-treated Zr-2.5 Nb pressure tubes reportedly show linear increase in in-reactor dimensional change with fast neutron fluence and appear to be showing better resistance to in-reactor dimensional changes (figure 15 of Coleman et al (2007)).…”

“…Cold worked and stress relieved Zr-2.5 Nb pressure tubes reportedly show low inreactor dimensional changes at lower fluences and shows accelerated behaviour at larger fluences (Coleman et al 2007). On the other hand, heat-treated Zr-2.5 Nb pressure tubes reportedly show linear increase in in-reactor dimensional change with fast neutron fluence and appear to be showing better resistance to in-reactor dimensional changes (figure 15 of Coleman et al (2007)). Irradiation growth, thermal creep and irradiation enhanced creep are the main in-reactor deformation modes for Zr-2.5 Nb alloy pressure tubes and are reportedly influenced by intrinsic parameters such as alloy chemistry, grain-size and its aspect ratio, texture, dislocation type and its density and extrinsic parameters such as temperature, operating stress and fluence (Coleman et al 2007).…”

“…Although pressure tubes of this material served satisfactorily in reactor applications, it is now known that Zircaloy-2 suffers from high rates of hydrogen pick up and somewhat high rates of irradiation induced creep. Though internationally Zr-2.5 Nb alloy has been tried in two different conditions: conventional cold worked and stress relieved and heat treated (quenched and aged) state (Coleman et al 2007), in India and Canada due to ease of fabrication and economic reasons Zr-2.5 Nb alloy is used as pressure tube in cold worked and stress relieved condition. The heat-treatment for the quenched and aged condition is carried out almost at the last stage after which very little cold work is imparted.…”

“…A literature survey (Coleman et al 2007) of the manufacturing practices of the Zr-2.5 Nb alloy pressure tubes revealed that this alloy is used in four metallurgical conditions, viz., cold worked and annealed (Russia); cold worked and stress-relieved (Canada, India, South Korea); and water quenched and aged (Russia, Japan) and Gas quenched and aged (Russia). The cold worked and annealed tubes and gas quenched and aged tubes have much lower strength and do not meet the short term strength requirement of the pressure tubes for Indian heavy water reactors.…”

“…On the other hand, heat-treated Zr-2.5 Nb pressure tubes reportedly show linear increase in in-reactor dimensional change with fast neutron fluence and appear to be showing better resistance to in-reactor dimensional changes (figure 15 of Coleman et al (2007)). Irradiation growth, thermal creep and irradiation enhanced creep are the main in-reactor deformation modes for Zr-2.5 Nb alloy pressure tubes and are reportedly influenced by intrinsic parameters such as alloy chemistry, grain-size and its aspect ratio, texture, dislocation type and its density and extrinsic parameters such as temperature, operating stress and fluence (Coleman et al 2007). Thus a need was felt to revisit the design limits, alloy chemistry and manufacturing practice of Zr-2.5 Nb alloy pressure tubes for existing fleet of PHWRs and new generation reactors such as PHWR700 and AHWRs.…”