H Tsai

Reference vanadium alloy V4Cr4Ti for fusion application

Vanadium alloys exhibit important advantages as a candidate structural material for fusion first-wall/blanket applications. These advantages include high temperature and high wall load capability, favorable safety and environmental features, resistance to irradiation damage, and alloys of interest are readily fabricable. A substantial data base has been developed on laboratory-scale heats of V-Ti, V-Cr-Ti and V-Ti-Si alloys before and after irradiation. Investigations in recent years have focused primarily on compositions of V-(0--15)Cr-(0--20)Ti (0--1)Si. Results from these investigations have provided a basis for identifying a V-4Cr-4Ti alloy as the US reference vanadium alloy for further development. Major results obtained on one production-scale heat and three laboratory heats with compositions of V-(4--5)Cr-(4--5)Ti are presented in this paper. Properties measured were input properties, tensile properties, creep, and radiation effects.

Fuel/cladding compatibility in U-19Pu-10Zr/HT9-clad fuel at elevated temperatures

Abstract Twenty-five elevated-temperature tests were conducted on irradiated U-19wt%Pu-10wt%Zr fuel clad with HT9 at the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory, Illinois. Samples with burnups as high as 11 at% were tested, and the cladding penetration rates all fall within the data band previously defined. Highlights of the test results are presented, as are discussions of the cladding penetration mechanism found in elevated-temperature testing of irradiated metallic fuels.

Performance of U-Pu-Zr fuel cast into zirconium molds☆

Abstract To investigate a means of eliminating quartz mold waste, U-3Zr and U-20.5Pu-3Zr fuel was injection cast into Zr tubes, or sheaths, and clad in 316SS. These elements and U-10Zr and U-19Pu-10Zr fuel elements, cast in the standard manner into disposable quartz molds, were irradiated in EBR-II to 2 at% burnup and removed for interim examination. Measurments of fuel column axial growth indicate that the Zr-sheathed fuel exhibited significantly less axial elongation than the standard-cast fuel (1.3 to 1.8% versus 4.9 to 8.1%). Fuel material extruded through the ends of the Zr sheaths, contacting the cladding in some cases. Transverse metallographic sections reveal cracks in the Zr sheath, indicating that the sheath is not a sufficient barrier between fuel and cladding to reduce fuel-cladding chemical interaction (FCCI). FCCI effects will be monitored as the elements attain higher burnup.

Modeling the behavior of metallic fast reactor fuels during extended transients

Abstract Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram systems fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements.

Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

Abstract Three furnace heating tests were conducted with irradiated, HT9-clad and U-19wt%Pu-10wt%Zr-alloy, EBR-II Mk-V-type fuel elements to evaluate the behavior that could be expected during a loss-of-flow event in the reactor. In general, very significant safety margins for cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results are presented, as are discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction that were found in these tests.

Tensile and impact properties of V–4Cr–4Ti alloy heats 832665 and 832864 ☆

Abstract Two large heats of V–4Cr–4Ti alloy were produced in the US in the past few years. The first, 832665, was a 500 kg heat procured by the US Department of Energy for basic fusion structural materials research. The second, 832864, was a 1300 kg heat procured by General Atomics for the DIII-D radiative divertor upgrade. Both heats were produced by Oremet-Wah Chang (previously Teledyne Wah Chang of Albany). Tensile properties up to 800°C and Charpy V-notch impact properties down to liquid nitrogen temperature were measured for both heats. The product forms tested for both heats were rolled sheets annealed at 1000°C for 1 h in vacuum. Testing results show the behavior of the two heats to be similar and the reduction of strengths with temperature to be insignificant up to at least 750°C. Ductility of both materials is good in the test temperature range. Impact properties for both heats are excellent – no brittle failures at temperatures above −150°C. Compared to the data for previous smaller laboratory heats of 15–50 kg, the results show that scale-up of vanadium alloy ingot production to sizes useful for reactor blanket design can be successfully achieved as long as reasonable process control is implemented (H. Tsai, et al., Fusion Materials Semiannual Progress Report for Period Ending 30th June 1998, DOE/ER-0313/24, p. 3; H. Tsai, et al., Fusion Materials Semiannual Progress Report for Period Ending 31st December 1998, DOE/ER-0313/25, p. 3).

Effects of low-temperature neutron irradiation on mechanical properties of vanadium-base alloys

Vanadium-base alloys were irradiated in three experiments to determine the effects of low-temperature neutron irradiation on their mechanical properties. The properties studied were tensile, Charpy impact and irradiation creep. The result of this study showed the alloys tested incurred significant hardening and embrittlement due to the irradiation. Heat-to-heat variations appeared to have little effect on hardening and embrittlement. Creep strains were measured from the specimens and a preliminary data set on creep strain rate was generated.

Irradiation creep of vanadium-base alloys

A study of irradiation creep in vanadium-base alloys is underway with experiments in the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) in the US. Test specimens are thin-wall sealed tubes with internal pressure loading. The results from the initial ATR irradiation at low temperature (200-300 C) to a neutron damage level of 4.7 dpa show creep rates ranging from {approx}0 to 1.2 x 10{sup {minus}5}/dpa/MPa for a 500-kg heat of V-4Cr-4Ti alloy. These rates were generally lower than reported from a previous experiment in BR-10. Because both the attained neutron damage levels and the creep strains were low in the present study, however, these creep rates should be regarded as only preliminary. Substantially more testing is required before a data base on irradiation creep of vanadium alloys can be developed and used with confidence.

Mechanical behavior and microstructural evolution of vanadium alloys irradiated in ATR-A1

Abstract An irradiation experiment has been done in the ATR-A1 to investigate irradiation behavior of vanadium alloys in the low temperature regime from 200°C to 300°C with damage levels of 3 to 4 dpa. In creep measurements, creep tubes of V–3Fe–4Ti–0.1Si with inner pressures up to 165 MPa did not rupture during irradiation. The effective strain rate of creep was below 0.2% dpa−1 and it showed the same tendency as V–4Cr–4Ti alloys. In Charpy impact tests, all specimens of V–4Cr–4Ti–0.1Si and V–3Fe–4Ti–0.1Si showed brittle behavior at room temperature and the DBTT increased to 60–150°C. The fracture surface showed cleavage. Tensile tests conducted both at room temperature and at the irradiation temperature showed significant irradiation hardening and brittle responses. TEM showed that high densities of tiny defect clusters were formed in V–Cr–Ti and V–Fe–Ti alloys. Precipitates could not be seen in specimens irradiated below 300°C, however, fine defect clusters are considered to be the origin of brittle behavior in V–Cr–Ti alloys irradiated at low temperatures.

Thermal Creep of Irradiated Zircaloy Cladding

As part of an effort to investigate spent-fuel behavior during dry-cask storage, thermal creep tests are being performed with defueled Zircaloy-4 cladding segments from two pressurized water reactors — Surry at ≈36 GWd/MTU burnup and H. B. Robinson at ≈67 GWd/MTU burnup, with corresponding fast (E > 1 MeV) fluence levels of 7×1025 and 14×1025 n/m2. The Surry rods are particularly relevant because they were stored in an inert-atmosphere (He) cask for 15 years. The Robinson rods were received after reactor discharge and pool storage. Commensurate with their high burnup, the Robinson cladding has significant waterside corrosion and hydrogen uptake. Test results to-date indicate good creep ductility for both claddings in the 360–400°C and 160–250 MPa (hoop-stress) regime. Partial recovery of radiation hardening may have occurred during the long tests at 400°C, which led to improved creep ductility. Creep-rate sensitivity is significant for stress and even more so for temperature. The higher hydrogen content in the Robinson material appears to have no detrimental effect on creep behavior at the test temperature. One Robinson sample, which ruptured in the weld region at 205°C during cooling from 400°C under stress (190 MPa), precipitated all visible hydrides in the radial direction.