Argonne National Laboratory-West Research Articles

EBR-II Superheater Duplex Tube Examination

After 30 yr of operation, the Experimental Breeder Reactor II (EBR-II) Superheater 710 at Argonne National Laboratory-West (now Idaho National Laboratory) was decommissioned. As part of its postservice examination, four duplex tube sections were removed and Charpy impact testing was performed to characterize the crack-arresting ability of nickel-bonded tube interfaces. A scanning electron microscopy (SEM) examination was also performed to characterize and identify changes in bond material microstructure. From room temperature to 400°C, all samples demonstrated ductility and crack-stopping ability similar to that exhibited by beginning-of-life samples. However, at a low temperature (–50°C), samples removed from the lower region of the superheater (near the sodium inlet) failed while those from the upper region (near the sodium outlet) did not. SEM analysis revealed that all the tube-tube interfaces showed evidence of iron diffusion into the nickel braze, which resulted in the formation of a multiphase diffusion structure. Yet, significant void formation was only observed in the bond layer of the tubes removed from the lower region. This may be due to a change in the crystal microstructure of one of the phases within the bond layer that occurs in the 350 to 450°C temperature range, which results in a lower density and the formation of porosity. Apparently, only the samples from the higher-temperature region were exposed to this transition temperature, and the resulting large voids that developed acted as stress concentrators that led to low-temperature embrittlement and failure of the Charpy impact specimens.

Plutonium-238 Alpha-Decay Damage Study of A Glass-Bonded Sodalite Ceramic Waste Form

Abstract An accelerated alpha-decay damage study of a glass-bonded sodalite ceramic waste form has been completed recently. The study was designed to investigate the physical and chemical durability of the waste form after exposure to 238Pu alpha decay. The alpha-decay dose at the end of the four year study was approximately 1.0 × 1018 decays/gram of material. The ceramic waste form (CWF) was developed to immobilize fission products accumulated during the treatment of spent nuclear fuel from the Experimental Breeder Reactor II performed at Argonne National Laboratory-West in Idaho. Small quantities of actinide elements are also found in the waste form. The CWF is currently undergoing qualification testing and characterization for acceptance by the Office of Civilian Radioactive Waste Management for geologic disposal. Methods used to monitor the 238Pu-loaded CWF material in this study included: immersion density determination to measure possible macroscopic swelling, chemical durability by leach testing, microstructural analysis by scanning and transmission electron microscopy, and phase composition and stability by powder X-ray and electron diffraction. After four years of exposure to alpha decay, the investigation observed little alteration to the CWF. Specifically, the 238Pu-loaded CWF maintained its physical integrity, the density remained constant, no cracking or phase debonding was observed by microscopy, and the material's chemical durability did not change significantly over the duration of the study. The only significant measured change was an increase of the unit-cell lattice parameters of the plutonium oxide and sodalite phases of the material, but these were very small and did not lead to any loss of waste form durability.

Determination of [formula omitted] in activated metal waste

The inherent radioactivity and potential toxicity of nuclear waste provide challenges to persons responsible for its characterization. Specifically, this paper describes the development of an analytical method quantitatively to measure 14C in radioactive waste. A separation from other beta-emitting radionuclides is first required before the weak beta radiation of 14C can be counted using liquid scintillation. Hence, this paper discusses the development of a modified, remote carbon analyzer which has provided a means to perform this separation. This modified technique is much easier than the difficult and time consuming techniques traditionally used for determining 14C concentrations. The carbon analyzer (LECO® HF-400/IR-412 carbon analyzer) that was modified for remote operation in the Argonne National Laboratory — West (ANL-W) Analytical Laboratory Hot Cells (ALHC) has been used to separate and isolate 14C from other beta-emitting radionuclides in samples of different matrices. These include irradiated beryllium from the Advanced Test Reactor (ATR) located on the Idaho National Environmental and Engineering Laboratory (INEEL) and samples of stainless steel that were irradiated in the Experimental Breeder Reactor II.

A Description of the Ceramic Waste Form Production Process from the Demonstration Phase of the Electrometallurgical Treatment of EBR-II Spent Fuel

The electrometallurgical treatment (EMT) process has been designed and developed for stabilizing sodium-bonded, metallic fuel into two high-level waste forms. This process has recently been successfully demonstrated with irradiated EBR-II fuel at Argonne National Laboratory-West. Part of the EMT process is to immobilize fission-product-bearing waste salt, which results from electrorefining, in a ceramic waste form—a glass-bonded sodalite. The sodalite is formed by hot isostatically pressing salt-loaded zeolite at temperatures up to 850°C and pressures up to 100 MPa. The specific unit operations that comprise ceramic waste production include steps for salt grinding, zeolite drying, blending salt and zeolite and glass frit in a v-blender, and consolidating the powders in a hot isostatic press. The results of testing these unit operations with irradiated salt from the EMT demonstration are summarized and include some preliminary characterization of the final irradiated ceramic waste form created by this process.

Morphologies of uranium and uranium–zirconium electrodeposits

Uranium and uranium–zirconium electrodeposits produced in the Fuel Cycle Facility at Argonne National Laboratory-West were examined using standard metallurgical techniques. Substantial differences in the morphologies of the two types of deposits were observed. Samples from pure uranium deposits were comprised of chains of uranium crystals with a characteristic rhomboidal shape, while the morphologies of samples from deposits containing zirconium in excess of approximately 0.5 wt% showed more polycrystalline features. Zirconium was found to be present as a second, zirconium metal phase at or very near the uranium–zirconium dendrite surfaces. Higher collection efficiencies and total deposit weights were observed for the uranium–zirconium deposits; this performance increase is likely a result of better mechanical properties exhibited by the uranium–zirconium dendrite morphology.

Hot demonstrations of nuclear-waste processing technologies

Several types of nuclear-waste-treatment technologies are currently being demonstrated at Argonne National Laboratory-West, ranging from complex, high-level waste-form production to a relatively simple encapsulation technique for the immobilization of mixed hazardous/radioactive waste. Typically, the demonstrations involve a cooperative arrangement among the organizations responsible for developing the processes, operating the facilities, and conducting the test programs. By summarizing the common approach taken for demonstrations of three very different waste-treatment processes, this article illustrates the necessity of process demonstrations using typical radioactive waste streams in the development of industrially viable waste-treatment technologies.

Criticality Safety Strategy and Analysis Summary for the Fuel Cycle Facility Electrorefiner at Argonne National Laboratory West

As part of the termination activities of Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory (ANL) West, the spent metallic fuel from EBR-II will be treated in the fuel cycle facility (FCF). A key component of the spent-fuel treatment process in the FCF is the electrorefiner (ER) in which the actinide metals are separated from the active metal fission products and the reactive bond sodium. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt, and refined uranium or uranium /plutonium products are deposited at cathodes. The criticality safety strategy and analysis for the ANL West FCF ER is summarized. The FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. To show criticality safety for the FCFER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOEs) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOEs, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that will verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCFER operations were found to be safe with respect to criticality.

Thermal/Hydrological Modeling of the Radioactive Scrap and Waste Facility (RSWF) with the Tough2 (Transport of Unsaturated Groundwater and Heat) Code

ABSTRACTThermal/hydrological modeling of the Radioactive Scrap and Waste Facility (RSWF) has been completed with the TOUGH2 (Transport of Unsaturated Groundwater and Heat) Code.[1] The RSWF will be utilized as an interim storage facility for ceramic and metallic waste forms developed from the electrometallurgical treatment of spent nuclear fuel. The RSWF is an array of 1,350 carbon steel liners located at grade near Argonne National Laboratory-West on the Idaho National Engineering Laboratory (INEL).The primary driving force for this modeling research was to assess thermal capacity limits for RSWF liners so that heat generating materials can be safely stored. Maximum wasteform temperatures will be governed by both the amount of heat the system can dissipate and the orientation and characteristics of the wasteform. The focus of this report is on the amount of heat the interim storage system can safely dissipate. The effect of the temporal variation of soil moisture on the performance of the RSWF is assessed. The facility was analyzed to determine the maximum allowable thermal loading of the RSWF liners.

Measurement of fission fragment induced fluorescence in xenon

A research program was conducted at Argonne National Laboratory-West (ANL-W) to measure the 172 nm fluorescence in xenon induced by fission fragments. Additionally, the survivability of uranium dioxide coatings under high fission power conditions was investigated, and the fission fragment escape energy from coatings of different thicknesses was directly measured. A definite rearrangement of the uranium dioxide coating was observed after high power irradiations, but this rearrangement did not affect the fission fragment energy escape fraction. The measurement of fluorescence induced by fission fragments was done in the Neutron Radiography Reactor (NRAD); this measurement was complicated by the presence of the gamma field and by gas impurity effects. Compensations for gamma and impurity effects were made during the program. Additionally, signal degradations during system pressure changes were observed; however, the signal recovered after several minutes, so this phenomenon did not adversely affect the measurements. The efficiencies of fission fragment and gamma induced fluorescence were measured to be 0.37 and 0.44 respectively.

Hot cell based time-of-flight mass spectrometry

A time-of-flight mass spectrometer has been designed and is presently being installed within a radiological controlled hot cell at Argonne National Laboratory-West. Direct solid sampling is performed by laser ablation or glow discharge sputtering followed by supersonic expansion into the TOFMS source chamber. Once the atom/ion beam enters the source chamber, enhanced ionization can be accomplished by laser ionization or electron impact. An assortment of samples may be analyzed ranging from irradiated nuclear fuel elements and cladding materials to eutectic salts, radioactive waste materials and environmental samples.