Postirradiation Ductility Demonstration Tests of Weapons-Derived MOX Fuel Cladding

Abstract

The U.S. Department of Energy (DOE) Fissile Materials Disposition Program (FMDP) is pursuing reactor irradiation of mixed uranium-plutonium oxide (MOX) fuel for disposal of surplus weapons-usable plutonium. Since most of the MOX fuel utilization experience has been with reactor-grade plutonium, it is desired to demonstrate that the unique properties of the surplus weapons-derived or weapons-grade (WG) plutonium do not compromise the applicability of this MOX experience base. A related question to be addressed for weapons-derived MOX fuel is that of ductility loss of the cladding. While irradiation induced loss of ductility has long been known and quantified for many cladding materials, the potential synergistic effects of irradiation and the unique constituents (i.e., gallium) of weapons-derived MOX fuel are not known. As part of an extensive fuel qualification research program conducted by Oak Ridge National Laboratory (ORNL), a new test method was developed and validated to measure the room temperature ductility and hoop tensile properties of MOX fuel cladding. The cladding material is a zirconium alloy designated as Zr-4 manufactured by Sandvick Corporation. This paper is a summary of the test method developed and of demonstration test results obtained for MOX cladding irradiated to 21 GWd/MT [7 × 1020 n/cm2 (E > 1 MeV)].

The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

“…11 are listed in Table 2. Unirriadiated M5 material properties are cited from Morris et al 2012 [5], Chapter 8 "Cladding mechanical testing." Fig.…”

“…The radial expansion of the specimen is used to calculate the circumferential strain accrued during the test. A general procedure has been developed to determine the circumferential stress (σ CIR ) in the expanded ring specimen [5]. The circumferential stress is calculated using a scaling parameter called a Г-factor (Г), the load (P), and the specimen dimensions as shown below.…”

Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding

“…Additional information regarding cleaning and cutting of cladding are provided elsewhere. 3,11 Polyurethane plugs with a Shore A hardness of 95 were machined with a diameter of 8.3 mm (0.33 in.) and a length of 7.5 mm (0.295 in.)…”

“…36 is a plot of 0.2% strain offset, and the intersection of this plot with stress has been used for estimating yield strength in previous reports. 3,11 For the unirradiated Zircaloy-4 data plotted in Fig. 36, the yield strength is 670 MPa.…”

“…2 The second phase covered ductility testing of irradiated MOX test fuel cladding for fast neutron fluences over the range 0.27-1.40 × 10 21 neutrons/cm 2 (E > 1 MeV). 3 The third phase extended the range of neutron fluences to 0.27-1.7 × 10 21 neutrons/cm 2 (E > 1 MeV). This report is a comprehensive presentation of all irradiated cladding ductility tests.…”