On the Embrittlement of Zircaloy-4 Under RIA-Relevant Conditions

Abstract

ABSTRACT:The extended use of Zircaloy cladding in light water reactors degrades its mechanical properties by a combination of irradiation embrittlement, coolant-side oxidation, hydrogen pickup, and hydride formation. The hydrides are usually concentrated in the form of a dense layer or rim near the cooler outer surface of the cladding. Utilizing plane-strain ringstretch tests to approximate the loading path in a reactivity-initiated accident (RIA) transient, we examined the influence of a hydride rim on the fracture behavior of unirradiated Zircaloy-4 cladding at room temperature and 300°C. Failure is sensitive to hydride-rim thickness such that cladding tubes with a hydride-rim thickness >100 µm (≈700 wppm total hydrogen) exhibit brittle behavior, while those with a thickness <90 µm (≈600 wppm) remain ductile. The mechanism of failure is identified as strain-induced crack initiation within the hydride rim and failure within the uncracked ligament due to either a shear instability or damage-induced fracture. We also report some preliminary results of the uniaxial tensile behavior of low-Sn Zircaloy-4 cladding tubes in a cold-worked, stress-relieved condition in the transverse (hoop) direction at strain rates of 0.001/s and 0.2/s and temperatures of 26-400°C.

An elasto-plastic fracture mechanics based model for assessment of hydride embrittlement in zircaloy cladding tubes

“…One important question is which criterion one should use for the crack growth. It can be assumed that the crack growth occurs by successive tearing of ligaments between fractured hydrides as shown in for instance [8,31], but the concept of J-controlled fracture implies that J completely characterizes the crack tip situation. When the plastic zone is large compared to a characteristic length, such as the crack ligament, and the shape of the plastic zone depends on the crack geometry then one should not expect that a simple 1-parameter such as J is an accurate parameter for fracture.…”

“…When the spent fuel is removed from the reactor and put into wet storage the hydrogen precipitates as hydrides. The hydrides often appear in more concentrated form as a hydride rim close to the surface and as individual elongated hydrides aligned primarily in the circumferential direction due to the metallurgical texture of the tube [8]. The internal pressure at room temperature is typically 3-7 MPa [9].…”

“…The internal pressure at room temperature is typically 3-7 MPa [9]. When the spent fuel is transferred from wet to dry storage the temperature increases to 400°C, the hydrides dissolve and the pressure increases (6)(7)(8)(9)(10)(11)(12)(13)(14)(15)(16). The hydrides are then reprecitated as partially radial hydrides as the temperature drops.…”

Behaviors of Nuclear Fuel Cladding During RIA

“…Accordingly, it was found that the high burn-up Zircaloy-4 cladding becomes very brittle even at the high temperatures achievable during a designbasis accident (Daum et al, 2002). Fig.…”

Ductile-to-Brittle Transition Temperature for High Burnup Zircaloy-4 and ZIRLO (TM) Cladding Alloys Exposed to Simulated Drying Storage Conditions.

“…For fueled cladding, the fuel resists decreases in cladding inner diameter by applying a secondary radial load to the cladding. Ductility of homogeneous metals tends to decrease as the biaxiality ratio (σ θ /σ z ) → 1 [22] and may decrease at higher displacement-rate loading. As such, use of offset strain determined just prior to development of a 100% wall crack due to RCT loading would lead to overly optimistic results.…”