Delayed hydride cracking (DHC) has been responsible for cracking in zirconium alloy pressure tubes and fuel cladding and is a concern for spent fuel storage. For cracking to start, sufficient hydrogen must be present for hydride to form at a flaw tip and the local tensile stress must be sufficiently large to crack the hydride (a crack will not extend if the threshold in the stress intensity factor, K_IH, is not exceeded. A high-temperature limit exists when the yield stress of the cladding alloy becomes too low to crack the hydride. In this paper we describe measurements of K_IH and the crack growth rate, V, in unirradiated Zircaloy-4 fuel cladding containing approximately 130 ppm hydrogen in the cold-worked stress–relieved condition representing pressurized water reactors (PWRs) and pressurized heavy-water (PHWR) reactors. Four methods are used to evaluate K_IH. The test specimen and fixture used in these methods was the pin-loading tension configuration. The test temperature ranged from 227 to 315°C. The mean value of K_IH below 280°C had little temperature dependence; it was about 5.5 MPa√m in the PWR cladding and slightly higher at 7 MPa√m in the PHWR material. At higher test temperatures, K_IH increased dramatically to more than 12 MPa√m, whereas the crack growth rate declined toward zero. This behavior suggests that unirradiated Zircaloy-4 fuel cladding is immune from DHC above about 320°C; this temperature may be increased to 360°C by irradiation. The implications for spent fuel storage are that during early storage when the temperatures are high, any flaw will not extend by DHC, whereas at low temperatures, after many years of storage, flaws would have to be very large, approaching through wall, before being extended by DHC. To date, spent nuclear fuel is not known to have failed by DHC during storage, confirming the inference.