Jack D. Galloway

Fuel-Cladding Chemical Interaction in U-Pu-Zr Metallic Fuels: A Critical Review

Abstract Fuel-cladding chemical interaction (FCCI) is a phenomenon that occurs at the fuel-cladding interface during the irradiation of U-Zr and U-Pu-Zr metallic nuclear fuel and stainless steel cladding. The inter-diffusion zone that develops places both the fuel and cladding at risk through the reduction in cladding strength and the formation of a (U,Pu)/Fe eutectic in the fuel. Due to the impact FCCI has on limiting fuel pin burnup, there is a need for better understanding of the governing FCCI mechanisms in order to make accurate predictions using fuel-performance codes. By performing a critical review of previous work, the physics of FCCI can be separated into individual phenomena so that targeted models can be developed for each. Through examination of experiments conducted both in- and out-of-reactor, the behavior of lanthanides provides a natural separation of models by tracking their behavior through (1) production and transport in the fuel to the clad, (2) interaction with macroscopic changes in fuel topography including cracking and swelling, and finally (3) inter-diffusion at the fuel-cladding interface. Informed by past experience, phenomenological models can be built for each separate effect and subsequently combined in an integral fuel-performance simulation. Prototypical simulation approaches at each level have been included, as well as suggestions for several experiments to help bolster the understanding of irradiated fuel. A robust and predictive FCCI model will provide fuel-performance codes with the ability to predict clad failure and/or fuel eutectic melting. Armed with this information, advanced concepts such as palladium doped fuel, ODS steels, or mitigating reactor designs may be able to reduce FCCI enough to extend fuel burnup beyond its current limits, potentially boosting safety margins and reducing cost through higher fuel utilization.

Accident-Tolerant-Fuel Performance Analysis of APMT Steel Clad/UO 2 Fuel and APMT Steel Clad/UN-U 3 Si 5 Fuel Concepts

Abstract While Zircaloy-based claddings have been the workhorse for the nuclear power industry for decades, they have also demonstrated problems, particularly regarding accident scenarios. Work has been performed to assess the viability of stainless steel–based cladding in traditional light water reactors. This paper assesses the reactivity penalty of moving to stainless steel cladding using Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify gains or losses in structural integrity when moving to thinner, stainless steel claddings. Thermal and irradiation creep, along with fission gas swelling, thermal swelling, and fuel relocation, are accounted for in the models for both Zircaloy and stainless steel claddings. Additional models for the lower-oxidation stainless steel APMT are also invoked where available, with irradiation data for HT9 used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied toward cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided.

Salt-Cooled Modular Innovative Thorium Heavy Water-Moderated Reactor System

Abstract This paper describes a new reactor concept: the Salt-cooled Modular Innovative THorium HEavy water-moderated Reactor System (SMITHERS), which addresses the goals of (a) evolving deployment needs, (b) increasing overall fuel burnup, (c) reducing proliferation risk, and (d) providing high-efficiency power generation. The reactor is modular and thus scalable from a few to hundreds of megawatts(thermal). The concept further burns used fuel from light water reactors (LWRs) without aqueous separations, reducing costs and proliferation pathways relative to current reprocessing plants. The additional burning of LWR fuel reduces proliferation risk by reducing global inventories of plutonium from used fuel in a way that does not isolate weapons-useable material and that increases the amount of power produced per ton of mined uranium. Improved fuel utilization through the potential use of thorium provides cost benefits by increasing neutron economy and enabling operation at higher efficiencies. Neutron economy is increased by using the lower neutron energies associated with large quantities of heavy water moderation and/or thorium for innovative reactor control and constant long-term power generation (i.e., sustainability). Finally, the proposed reactor also generates high-temperature coolant discharge in the form of liquid salt without coolant pressurization for external process heat applications such as oil extraction. Salt offers significant improvement over existing coolants such as light water and heavy water, which require pressurization to operate at high temperatures, adding to the cost and complexity of reactor operation. SMITHERS designs discussed in this paper either burned a full core of used fuel, ThO2 with 1.2 wt% PuO2 or other fissile material, or a combination of the two.

Uncertainty Quantification for New Approaches to Spent Fuel Assay

Abstract Estimating plutonium (Pu) mass in spent nuclear fuel assemblies (SFAs) helps inspectors ensure that no Pu is diverted. Therefore, nondestructive assay (NDA) methods are being developed to assay Pu mass in SFAs. Uncertainty quantification is an important task in most assay methods, and particularly for SFA assay. A computer model (MCNPX) is being used to predict isotope masses and the spatial distribution of masses in virtual SFAs for 64 combinations of initial fuel enrichment (IE), fuel utilization [burnup (BU)], and cooling time (CT) values. Additional MCNPX modeling for the same 64 virtual SFAs provided the expected detector responses (DRs) for several NDA techniques such as the passive neutron albedo reactivity method and the 252Cf interrogation with prompt neutrons method. A previous paper describes one uncertainty quantification approach involving Monte Carlo (MC) simulation using individually any of six new NDA options together with IE, BU, and CT. This paper provides an interpretation of the MC approach that is suited for a numerical Bayesian alternative, separately assesses the impact of MCNPX interpolation error, and compares several options to use subsets of IE, BU, CT, and one DR.

Constituent Redistribution in U-Zr Metallic Fuel Using the Advanced Fuel Performance Code BISON

Previous work done by Galloway, et. al. [1] on EBR-II ternary (U-Pu-Zr) fuel constituent redistribution yielded accurate simulation data for the limited data sets of Zr redistribution. The data sets included EPMA scans of two different irradiated rods. First, T179 which was irradiated to 1.9 at% burnup was analyzed. Second, DP16 which was irradiated to 11 at% burnup was analyzed. One set of parameters that most accurately represented the zirconium profiles for both experiments was determined. Since the binary fuel (U-Zr) has previously been used as the driver fuel for sodium fast reactors (SFR) as well as being the likely driver fuel if a new SFR is constructed, this same process has been initiated on the binary fuel form. From limited binary EPMA scans as well as other fuel characterization techniques, it has been observed that zirconium redistribution also occurs in the binary fuel, albeit at a reduced rate compared to observation in the ternary fuel, as noted by Kim et. al. in [2]. While the rate of redistribution has been observed to be slower, numerous metallographs of U-Zr fuel, such as the one shown in Figure 1, show distinct zone formations.