Uranium-zirconium (U-Zr) alloys are candidate fuels���for fast reactor designs under development���by���private���industry���and the���U.S.���Department of Energy.���However, U-Zr fuels are not currently qualified for use within these reactors. Fuel qualification is a���costly���process that typically takes over 20 years from the conceptualization of the fuel type���followed by prolific neutron irradiations and post-irradiation examinations. Recently,���co-PI���Beausoleil et al. proposed a new paradigm to���dramatically reduce the time required for���the fuel qualification process using accelerated neutron irradiations���and integrated���advanced modeling.���An example of the accelerated fuel testing proposed is the���Fission Accelerated Steady-state Test (FAST), an integral effects experiment,���which���can reach higher burnups at���reduced times compared to traditional in-pile testing. The basis of this���concept is that the FAST irradiation���maintains an equivalent linear heat generation rate���compared���to���prototypic fuel (e.g., EBR-II pins) while radially scaling the dimensions to obtain���a target burnup more rapidly.��� Historically, a���major limiting factor for the burnup of U-Zr alloys is fuel-cladding chemical interaction (FCCI), shown in���Fig 1(a),���in���which���fission products and���U���have deleterious interactions with the cladding alloy constituents (e.g., U-Fe eutectics or���an���Fe-Nd intermetallic). FCCI is a critical area of research for sodium fast reactor research���since���it is a primary contributor to cladding���wastage���(indicated as���FCCI���in Fig 1(a))���and eventual cladding failure���(Fig. 1(b)). For U-10Zr (10 weight percent Zr)���alloy fuel, the onset of lanthanide-based FCCI is typically observed at���~10%���FIMA (fissions of initial metal atoms) burnup.���Therefore, cladding failure in EBR-II fuel was typically observed at burnups of 15-20%���FIMA with very few experiments reaching 20%���FIMA.���To address this early failure, cladding liners, serving as a diffusion barrier between the fuel and cladding, have���recently���been proposed to mitigate FCCI.��� By combining the���PIE of the���FAST experiments���using���advanced characterization techniques and���modern���modeling tools, we will���test���the following���hypotheses:���1) Are the FAST integral experiments suitable for accelerating neutron irradiations to replicate similar irradiation-induced phenomena observed in historical, prototypical in-pile irradiations?���2) Do Zr liners effectively mitigate���FCCI for U-Zr? 3) Can���the diffusion of lanthanides���be modeled���within���the fuel, Zr liner, and FCCI regions to improve the predictions of���FCCI evolution as a function of burnup?���

TRISO particles are frequently advertised as ���functional containment��� for reactors due to their high retention of the fission products created in the uranium kernel at their core. However, statistically some of these TRISO particles will fail. This project proposes a multi-step investigation combining expertise at NCSU and EPRI on the atomistic and mesoscale to understand the transport of fission products, determine associated stresses throughout the lifetime of the particle in individual layers, and how the associated stresses can lead to cracks and any other failure modes. This will help develop an understanding of the way in which fission products may be released into the graphite matrix containing the TRISO particles and provide insight into potential advantages gained from advanced TRISO-based fuel particle designs.

A monolithic UMo fuel with low enriched fuel is proposed in place of the high enrichment UMo fuels used in research reactors. Qualification of this new type of fuel requires the fuel to maintain stable and predictable behavior throughout its lifetime in-reactor. Mechanistic fuel models are being developed that both correspond to existing experimental data on fuel swelling and can be applied to irradiation conditions beyond the experimental scope. In order to develop such mechanistic models, accurate fuel property data is required and must be obtained from either experiments or lower length scale modeling methodologies. One such parameter of critical importance is the species diffusion in UMo, with special emphasis on fission gas diffusion. There exists experimental data on high temperature, intrinsic diffusion of U and Mo. The Microstructural Modeling Working Group formed under the USHPRR program atomistically calculated the radiation driven diffusion coefficients. However, no such data on radiation enhanced diffusion in UMo alloys exists. In support of the USHPRR program, this subcontract supports the calculation of radiation enhanced diffusion of U, Mo and Xe in UMo fuels for application to fission gas swelling mechanistic models.

A significant knowledge gap exists in the data for the fundamental properties relevant to fuels and coolants for molten salt reactors (MSRs) that needs to be addressed in order to expedite the technical readiness level of the MSR design concepts. The US-DOE has identified the need to better understand, predict, and optimize the physical properties and thermochemical behavior of molten salts. In connection with INL LDRD XXXX, this subcontract outlines proposed scope for ab initio molecular dynamics (AIMD) simulations to be conducted in collaboration with INL, to support the development of an experimentally validated computational framework to extend the limited experimental data of fundamental thermophysical properties for a variety of molten salts to cover a broader range of composition and temperature.

Metallic fuels such as U-Zr and U-Pu-Zr are being proposed for certain new reactor designs, such as microreactors and the Versatile Test Reactor (VTR). Idaho National Laboratory (INL) can support metallic fuel development via fuel material research and the Bison fuel performance code. The existing metallic fuel performance models are empirical and do not match existing data very well. To accelerate materials discovery fuel design timelines and to improve fuel performance models, mechanistic (physics-based) models of fuel swelling, fission gas venting and fuel creep are necessary. The ������������-uranium phase exists in U-Zr and U-Pu-Zr fuel and significantly contributes to fuel behavior, but many fundamental materials properties and mechanisms of ������������-uranium are lacking. In support of INL metallic fuel performance modeling development, and under the INL LDRD 20A44-121, this subcontract provides for atomistic modeling to understand the mechanisms of irradiation damage in ������������-uranium and the effect of interfaces. These atomistic modeling simulations will be performed in collaboration with an experimental characterization campaign to construct and inform a mesoscale evolution model of defect migration and evolution and the impact on microstructure.