Jangyul Park

Evaluation of stainless steel–zirconium alloys as high-level nuclear waste forms

Abstract Stainless steel–zirconium (SS–Zr) alloys have been developed for the consolidation and disposal of waste stainless steel, zirconium, and noble metal fission products such as Nb, Mo, Tc, Ru, Pd, and Ag recovered from spent nuclear fuel assemblies. These remnant waste metals are left behind following electrometallurgical treatment, a molten salt-based process being demonstrated by Argonne National Laboratory. Two SS–Zr compositions have been selected as baseline waste form alloys: (a) stainless steel–15 wt% zirconium (SS–15Zr) for stainless steel-clad fuels and (b) zirconium–8 wt% stainless steel (Zr–8SS) for Zircaloy-clad fuels. Simulated waste form alloys were prepared and tested to characterize the metallurgy of SS–15Zr and Zr–8SS and to evaluate their physical properties and corrosion resistance. Both SS–15Zr and Zr–8SS have multi-phase microstructures, are mechanically strong, and have thermophysical properties comparable to other metals. They also exhibit high resistance to corrosion in simulated groundwater as determined by immersion, electrochemical, and vapor hydration tests. Taken together, the microstructure, physical property, and corrosion resistance data indicate that SS–15Zr and Zr–8SS are viable materials as high-level waste forms.

Microstructure and phase identification in type 304 stainless Steel-Zirconium alloys

Stainless steel-zirconium alloys have been developed at Argonne National Laboratory to contain radioactive metal isotopes isolated from spent nuclear fuel. This article discusses the various phases that are formed in as-cast alloys of type 304 stainless steel and zirconium that contain up to 92 wt pct Zr. Microstructural characterization was performed by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS), and crystal structure information was obtained by X-ray diffraction. Type 304SS-Zr alloys with 5 and 10 wt pct Zr have a three-phase microstructure—austenite, ferrite, and the Laves intermetallic, Zr(Fe,Cr,Ni)2+x. whereas alloys with 15, 20, and 30 wt pct Zr contain only two phases—ferrite and Zr(Fe,Cr,Ni)2+x. Alloys with 45 to 67 wt pct Zr contain a mixture of Zr(Fe,Cr,Ni)2+x and Zr2(Ni,Fe), whereas alloys with 83 and 92 wt pct Zr contain three phases—α-Zr, Zr2(Ni,Fe), and Zr(Fe,Cr,Ni)2+x. Fe3Zr-type and Zr3Fe-type phases were not observed in the type 304SS-Zr alloys. The changes in alloy microstructure with zirconium content have been correlated to the Fe-Zr binary phase diagram.

BWR pipe crack remedies evaluation

Abstract This paper presents results on: (a) the influence of simulated BWR environments on the stress-corrosion-cracking (SCC) susceptibility of Types 304, 316NG, and 347 stainless steel (SS), (b) fracture-mechanics crack growth rate measurements on these materials and weld overlay specimens in different environments, and (c) residual stress measurements and metallographic evaluations of conventional pipe weldments treated by a mechanical-stress-improvement process (MSIP) as well as those produced by a narrow-gap welding, procedure. Crack initiation studies on Types 304 and 316NG SS under crevice and non-crevice conditions in 289°C water containing 0.25 ppm dissolved oxygen with low sulfate concentrations indicate that SCC initiates at low strains (3%) in the nuclear grade material. Crack growth measurements on fracture-mechanics-type specimens, under low-frequency cyclic loading, show that the Type 316NG steel cracks at a somewhat lower rate (−40%) than sensitized Type 304 SS in an impurity environment with 0.25 ppm dissolved oxygen; however, the latter material stops cracking when sulfate is removed from the water. Crack growth in both materials ceases under simulated hydrogen-water chemistry conditions (5 ppb oxygen) even with 100 ppb sulfate present in the water. An unexpected result was obtained in the test on a weld overlay specimen in the impurity environment, viz., the crack grew to the overlay interface at a nominal rate, branched at 90° in both directions, and then grew at a high rate (parallel to the nominal applied load). Residual stress measurements on MSIP-treated weldments and those produced by a narrow-gap welding procedure indicate that these techniques produce compressive stresses over most of the inner surface near the weld and heat-affected zones.

Stainless steel-zirconium alloy waste forms for metallic fission products and actinides during treatment of spent nuclear fuel

Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion.

BWR pipe crack and weld clad overlay studies

Leaks and cracks in the heat-affected zones of weldments in austenitic stainless steel piping in boiling water reactors (BWRs) due to intergranular stress corrosion cracking (IGSCC) have been observed since the mid-1960s. Since that time, cracking has continued to occur, and indication have been found in all parts of the recirculation system, including the largest diameter lines. Proposed solutions for the problem include procedures that produce a more favorable residual stress state on the inner surface, materials that are more resistant to stress corrosion cracking (SCC), and changes in the reactor environment that decrease the susceptibility to cracking. In addition to the evaluation of these remedies, it is also important to gain a better understanding of the weld overlay procedure, which is the most widely used short-term repair for flawed piping.

Manufacturing Stress Corrosion-Cracking Tube Specimens for Eddy Current Technique Evaluation

To detect degradation in steam generator (SG) tubes, periodic inspection using nondestructive examination techniques, such as an eddy current testing, is a common practice. Therefore, it is critical to evaluate and validate the reliability of the eddy current technique for ensuring the structural integrity of the SG tubes. The eddy current technique could be evaluated by comparing the data estimated by the eddy current with the destructive examination data of field cracks, which would be both costly and labor intensive. A viable alternative to pulled tube data is to manufacture crack specimens that closely represent actual field cracks in laboratory environments. A crack manufacturing method that can be conducted at room temperature and atmospheric pressure conditions is proposed. The method was applied to manufacture different types of stress corrosion cracking (SCC) specimens: axial outer-diameter (OD) SCC for straight tubes, circumferential ODSCC and primary water SCC (PWSCC) at hydraulic expansion transition regions, and axial PWSCC at the apex and tangential regions of U-bend tubes. To help the growth of SCC into the tube, corrosive chemicals (sodium tetrathionate) and tensile stress were applied. Eddy current and destructive examination data for SCC specimens were compared with the available field crack data to determine whether those SCC specimens are representative. It was determined that the proposed method could manufacture the representative crack specimens.

Evaluation of stainless steel pipe cracking: Causes and fixes☆

Abstract This paper discusses (1) studies of impurity effects on susceptibility to intergranular stress corrosion cracking (IGSCC), (2) intergranular crack growth rate measurements, (3) finite-element studies of the residual stresses produced by induction heating stress improvement (IHSI) and the addition of weld overlays to flawed piping, (4) leak-before-break analyses of piping with 360° part-through cracks, and (5) parametric studies on the effect of through-wall residual stresses on intergranular crack growth behavior in large diameter piping weldments. The studies on the effect of impurities on IGSCC of Type 304 stainless steel show a strong synergistic interaction between dissolved oxygen and impurity concentration of the water. Low carbon stainless steel (Type 316NG) appear resistant to IGSCC even in impurity environments. However, they can become susceptible to transgranular SCC with low levels of sulfate or chloride present in the environment. The finite-element calculations show that IHSI and the weld overlay produce compressive residual stresses on the inner surface, and that the stresses at the crack tip remain compressive under design loads at least for shallow cracks.

Manufacturing Stress Corrosion Cracking Tube Specimens for Eddy Current Technique Evaluation

To detect degradation in steam generator (SG) tubes, periodic inspection using non-destructive examination techniques, such as an eddy current testing, is a common practice. Therefore, it is critical to evaluate and validate the reliability of the eddy current technique for ensuring the structural integrity of the SG tubes. The eddy current technique could be evaluated by comparing the data estimated by the eddy current with the destructive examination data of field cracks, which would be both costly and labor intensive. A viable alternative to pulled tube data is to manufacture crack specimens that closely represent actual field cracks in laboratory environments. A crack manufacturing method that can be conducted at room temperature and atmospheric pressure conditions is proposed. The method was applied to manufacture different types of stress corrosion cracking (SCC) specimens: axial outer-diameter (OD) SCC for straight tubes, circumferential ODSCC and primary water SCC (PWSCC) at hydraulic expansion transition regions, axial PWSCC at the apex and tangential regions of U-bend tubes. To help the growth of SCC into the tube, corrosive chemicals (sodium tetrathionate) and tensile stress were applied. Eddy current and destructive examination data for SCC specimens were compared with the available field crack data to determine whether those SCC specimens are representative. It was determined that the proposed method could manufacture the representative crack specimens.Copyright

Time Dependent Leak Rate Change in SCC Degraded SG Tubes of PWRs

Primary water stress corrosion cracking (SCC) and an outside diameter SCC have occurred in the steam generator (SG) tubes of nuclear power plants in the republic of Korea and around the world. Although high corrosion resistant alloy 690 has been replacing the alloy 600 tubings, it is important to establish the repair criteria for the remaining degraded alloy 600 tubings to reassure regarding the reactors integrity, and still maintain the plugging ratio within the limits needed for its efficient operations. For assessment and management of the degradation, it is crucial to understand the initial leak behaviors and time dependent leak rate change from SCC flaws under a constant pressure. Stress corrosion cracked tube specimens were prepared at room temperature by the contact of sodium tetrathionate solution. The initial leak rate and time dependent leak rate were measured at different pressure ranges with time. Water pressure inside the tube was slowly increased in a step like manner with a designated holding time. The leak rate was calculated by dividing the amount of water by the time. A large open and long axial crack showed an increasing leak rate with time at a constant pressure, whereas small opened cracks did not show an increase in a time dependent leak rate. Under some pressures, the leak rate did not increase with the increase of pressure due to a tightness of circumferential cracks. Throughwall axial crack of 5 mm long may exhibit the leakage of action level 1 of the EPRI leakage guideline.Copyright

Effect of Lead on Passivation of Alloy 600 Surface

Abstract Effects of Pb on passivity of Alloy 600 surface were investigated in mild acidic aqueous solutions at 90°C using polarization, electrochemical impedance spectroscopy (EIS), Auger electron spectroscopy (AES) and X-ray photoelectron spectroscopy (XPS). The results indicate that Alloy 600 surface consists of an inner oxide layer and outer hydroxide layer, containing Cr3+ and Ni2+. Pb is incorporated into the surface film and increases the electronic conductivity of Cr oxide in the surface film. Oxidation of Ni is inhibited by the presence of Pb.