R.V. Strain

Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

Abstract Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium–molybdenum (U–Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4–10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235 U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel–matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U–10Mo composition. Both of the U–10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

Stabilization of Rocky Flats Pu-contaminated ash within chemically bonded phosphate ceramics

Abstract A feasibility study was conducted on the use of chemically bonded phosphate ceramics for stabilization of combustion residue of high transuranic (TRU) wastes. Using a matrix of magnesium potassium phosphate formed by the room-temperature reaction of MgO and KH 2 PO 4 solution, we made waste forms that contained 5 wt% Pu to satisfy the requirements of the Waste Isolation Pilot Plant. The waste forms were ceramics whose compression strength was twice that of conventional cement grout and whose connected porosity was ≈50% that of cement grout. Both surrogate and actual waste forms displayed high leaching resistance for both hazardous metals and Pu. Hydrogen generation resulting from the radiolytic decomposition of water and organic compounds present in the waste form did not appear to be a significant issue. Pu was present as PuO 2 that was physically microencapsulated in the matrix. In the process, pyrophoricity was removed and leaching resistance was enhanced. The high leaching resistance was due to the very low solubility of PuO 2 coupled with superior microencapsulation. As a result, the waste forms satisfied the current Safeguard Termination Limit requirement for storage of TRU combustion residues.

Analysis of constituent redistribution in the γ (bcc) U-Pu-Zr alloys under gradients of temperature and concentrations

Abstract Rods of a ternary alloy (71U–19Pu–10Zr by weight percent) were annealed under a temperature gradient of 220°C/cm for 41 days and examined for micro-structural development and compositional redistribution. An enrichment of Zr with concurrent depletion of U was observed within the γ (bcc) phase region on the hot-end side (T≅740°C). The experimental redistribution of the elements in the γ (bcc) phase was analyzed in the framework of multicomponent mass transport with due consideration of thermotransport and ternary diffusional interactions. Based on a new analysis involving an integration of interdiffusion fluxes in the diffusion zone, kinetic parameters related to the thermotransport and ternary interdiffusion were calculated for each component i over selected ranges of composition. The thermotransport coefficients of U, Pu, and Zr were in the approximate ratio of 1:2:−4.5 in the hot-end region. In addition, the interdiffusion flux contributions arising from the gradients of temperature and concentrations of U and Zr were estimated.

Fuel-sodium reaction product and its influence on breached mixed-oxide fuel pins

Abstract The formation and consequences of fuel-sodium reaction product (FSRP) in mixed-oxide fuel pins that were irradiated in EBR-II are described. These results indicated that the amount of FSRP that forms as a result of sodium entering a pin with failed cladding is limited by oxygen availability and sodium availability at a favorable fuel temperature. The formation of FSRP is limited to an outer band of the fuel during normal operating conditions. FSRP results in an effective increase in the smeared density of the fuel since the density of FSRP is about half that of the fuel. The FSRP also lowers the O/M of the fuel which reduces its thermal conductivity and increases its thermal expansion. Results from the irradiation tests indicate that the FSRP formation in-reactor is consistent with the characteristics found for sodium uranate and sodium uranoplutonate in the laboratory.

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.

Characteristics of Hydride Precipitation and Reorientation in Spent-Fuel Cladding

The morphology, number density, orientation, distribution, and crystallographic aspects of Zr hydrides in Zircaloy fuel cladding play important roles in fuel performance during all phases before and after discharge from the reactor, i.e., during normal operation, transient and accident situations in the reactor, temporary storage in a dry cask, and permanent storage in a waste repository. In the past, partly because of experimental difficulties, hydriding behavior in irradiated fuel cladding has been investigated mostly by optical microscopy (OM). In the present study, fundamental metallurgical and crystallographic characteristics of hydride precipitation and reorientation were investigated on the microscopic level by combined techniques of OM and transmission electron and scanning electron microscopy (TEM and SEM) of spent-fuel claddings discharged from several boiling and pressurized water reactors (BWRs and PWRs). Defueled sections of standard and Zr-lined Zircaloy-2 fuel claddings, irradiated to fluences of {approx}3.3 x 10{sup 21} n cm{sup {minus}2} and {approx}9.2 x 10{sup 21} n cm{sup {minus}2} (E > 1 MeV), respectively, were obtained from spent fuel rods discharged from two BWRs. Sections of standard and low-tin Zircaloy-4 claddings, irradiated to fluences of {approx}4.4 x 10{sup 21} n cm{sup {minus}2}, {approx}5.9 x 10{sup 21} n cm{sup {minus}2}, and {approx}9.6 x 10{sup 21} n cm{sup {minus}2} (E > 1 MeV) in three PWRs, were also obtained. Microstructural characteristics of hydrides were analyzed in as-irradiated condition and after gas-pressurization-burst or expanding-mandrel tests at 292-325 C in Ar for some of the spent-fuel claddings. Analyses were also conducted of hydride habit plane, morphology, and reorientation characteristics on unirradiated Zircaloy-4 cladding that contained dense radial hydrides. Reoriented hydrides in the slowly cooled unirradiated cladding were produced by expanding-mandrel loading.

Mechanical properties and microstructural characteristics of laser and electron-beam welds in V-4Cr-4Ti.

Mechanical properties and microstructural characteristics of laser and electron-beam welds of a 500-kg heat of V4Cr4Ti were investigated in as-welded condition and after postwelding heat treatment by impact testing, microhardness measurement, optical microscopy, X-ray diffraction, and transmission electron microscopy (TEM). Ductile-brittle transition temperatures of the laser and electron-beam welds were significantly higher than that of the base metal. However, excellent impact properties could be restored in both types of welds by postwelding annealing at 1000 C for 1 h in vacuum. Analysis by TEM revealed that annealed weld zones were characterized by extensive networks of fine V(C,O,N) precipitates, which clean away O, C, and N interstitial from the grain matrices. This process is accompanied by simultaneous annealing-out of the dense dislocations present in the weld zone. This finding could be useful in identifying an optimal welding procedure by controlling and adjusting the cooling rate of the weld zone by an innovative method to maximize the precipitation of V(C,O,N).