Creep curves for Diametral Creep of Zircaloy-2 Tubes in a Fast-Flux of 2.56 E13 n/cm2 s a t 258°C. .. . 35 9 Curves for Diametral Creep of Quenched, Cold-Drawn and Aged Zr-2. 5% Nb, Out-Of-Flux and in a Fast Neutron Flux (2.3 E13 n/cm2 sec) 1 Mev at 295°C.. .. .. 36 Curves for Diametral Creep of Cold Drawn Zr-2.5% Nb Out-Of-Flux and in a Fast Neutron Flux (1.9 E13 n/cm2 sec)

DISCLAIMERThis report was prepared as an accoun t of work sponsored b y an age ncy of the United States Gove rnm ent. Nei th er the United States Gove rnm en t nor any agency thereof, nor any of the ir em p lo yees, makes any wa r ranty, exp ress or implied, or assum es any legal li abi lity or respon sib i li ty for th e accuracy, completeness, or usefulness of an y i nformation , apparat us, pro d uct, or p ro cess disclosed, or represents t hat its use w o uld not inf ri nge pr ivat ely own ed rights . Reference herein t o any specifi c commercial product, proce ss, o r service by trade name, trade mark , ma nu facture r, o r oth erw ise, d oes no t n ecessa rily constitute or imply its end o rs ement, reco mmen dation , o r favori ng by th e United States Governm ent or an y age ncy th ereof . Th e views and o p inion s of authors expressed here in do no t necessa ri ly state or re flect t hose o f the Un ited States Government o r any age ncy t hereof. PA CIF IC N O RTHWE ST LA BO RATO RY op era ted by BA HE LLE for the UNI TE D STATES D EPARTMENT O F ENERG Y u nd er Contract DE-AC06-76RLO 1830 Pr inted in th e Uni te d Sta tes of Am er ica Availa ble from N ationa l Tec hni ca l Info rm ati on Se rvice U nited States D epa rtme nt of Co m me rce 5285 Po rt Roya l Road Sp ringfield . Vi rginia 22151 NTIS Price Codes Microfiche AOl ABSTRACTThis report presents the results of a study conducted by Pacific Northwest Laboratory (PNL) to determine the probable arrival condition of spent lightwater reactor (LWR) fuel after handling and interim storage in spent fuel storage pools(a) and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits(b) for storing, handl irig, and transporting unconsol idated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed.(a) Of the fuel bundles currently in storage, about 93% are in pools at reactor sites and about 7% are at independent pool facilities.

Fracture Toughness of Hydrogen Charged Zircaloy-2 Fracture Toughness, KI , of Hydrogen Charged Zircaloy-2 Irradiated ~t 280°C Fracture Toughness of Irradiated Zircaloy-2 Fracture Toughness of Cold-Worked Zircaloy-2 PRTR Pressure Tube Assembly Comparison of Effects of 100°C Irradiation on the Room Temperature Tensile Properties of Zircaloy-2 and Zircaloy-4 Engineering Stress-Strain Curves for Zr-2.5 wt% Nb Alloy Quenched from (a + a)-Phase and Aged 24 hr at 500°C Engineering Stress-Strain Curves for Zr-2.5 wt% Nb Alloy Quenched from a-Phase and Aged 24 hr at 500°C Engineering Stress-Strain Curves for Zr-2.5 wt% Nb Alloy Slow Cooled from (a + B)-Phase Engineering Stress-Strain Curves for Zr-2.

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Tubes of 2.5Nb zirconium alloy were fatigued in the tension-tension mode by cyclic internal pressures to cause axial crack growth and unstable fractures at room temperature. Pressure-cycle rates ranged from 400 to 3000 cph. Both the cold-worked and heat-treated conditions, before and after hydriding (200 to 300 ppm H2), were investigated. Exploratory tests were done to determine the effect of the axial length of the surface-stress-intensifying defect on fatigue-crack initiation, growth, and critical length at unstable fracture. From short (≤tube-wall thickness) length defects, fatigue-crack initiation and growth will occur at nominal peak hoop stresses equal to or less than the estimated endurance limit (∼25,000 psi). For fatigue cracks initiated at the outer diameter surface, the shape of the crack front is semicircular. The number of stress cycles required to initiate fatigue cracking at the surface is inversely proportional to the square of the surface-defect length. Crack-growth rates, for non-hydrided tubing are proportional to (ΔK)5, and to (ΔK)4 after hydriding. There was a tendency for the critical crack length at unstable fracture to decrease with an increase in the number of fatigue cycles required to initiate fatigue cracking. The fracture toughness of the heat-treated tubing before hydriding is less than for the cold-worked condition either before or after hydriding, and after hydriding the fracture toughness is further reduced.