Laboratory for Advanced Nuclear Nonproliferation and Safety

Fuel-clad thermo-mechanics

— of advanced solid-fueled reactor designs

Advanced nuclear reactor designs have gained traction over recent years due to their potential for increased thermodynamic efficiencies over traditional light-water reactors [1]. However, the performance of fuel-clad and structural materials impedes the realization of these concepts due to the combined effects of corrosion, radiation, and thermal stresses at high coolant temperatures (500+ °C) [2]. Thus, adequate understanding of the thermal-hydraulic and thermo-mechanical performance of novel, alternative fuel-clad materials is required. Though several computational tools and subcodes such as FRAPCON [3], BISON [4], and FRIPAC [5] accurately describe the behavior of nuclear fuels and their cladding materials, none serve a dedicated role in developing a full-core, thermal-hydraulic and thermo-mechanical profile of advanced reactors outfitted with alternative cladding materials. REX (Reactor Envelope Expansion), a novel reactor analysis tool [6], fills this need by relating full-core fission rates and depletion statistics from the Serpent 2 Monte Carlo neutron transport code [7] to reactor thermal-hydraulics and thermo-mechanics.

Developed for the analysis of advanced reactors outfitted with multi-metallic layered composite (MMLC) fuel-clad materials, REX performs core-wide, pin-specific thermal-hydraulic and thermo-mechanical calculations with the goal of identifying cladding materials that can improve reactor thermal performance. It iteratively pushes coolant inlet temperatures and dynamically modifies fuel pin geometries to force mechanical failure of a specified “original” cladding material. It then uses these fuel pin geometries as comparison criteria and tests the thermal-hydraulic and thermo-mechanical performance of alternative cladding materials.

MMLCs are an innovative reactor core material concept wherein a radiation-tolerant structural material is diffusion-bonded to corrosion-resistant-material layers. Expected to survive mechanical stresses and corrosion rates that would be deleterious for traditional nuclear fuel cladding materials, MMLCs are projected to permit increased coolant inlet temperatures and consequently boost reactor thermodynamic efficiencies. MMLCs are modeled in REX as standard materials (T91 steel in the case of the lead-cooled fast reactor (LFR)) [2] wrapped in unstressed corrosion-resistant layers that are isothermal with respect to the reactor coolant. As such, they are theoretical materials that conduct less heat, have a smaller stressable cross section area, but experience less corrosion and lower overall equivalent stresses.

References

  1. LOCATELLI, M. MANCINI, N. TODESCHINI, “Generation IV Nuclear Reactors: Current Status and Future Prospects”, Energy Policy, 61:1503–1520 (2013).
  2. P. SHORT, R. G. BALLINGER, H.E. HäNNINEN. “Corrosion Resistance of Alloys F91 and Fe–12Cr–2Si in Lead–Bismuth Eutectic up to 715°C” Journal of Nuclear Materials, 434, (1-3):259-281 (2013).
  3. GEELHOOD, W. LUSCHER, P. RAYNAUD, I. PORTER, “FRAPCON-4.0: A Computer Code for the Calculation of Steady-State, Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burnup”, (PNNL-19418, Vol.1 Rev.2):1.1–1.2, Pacific Northwest National Laboratory, (2015).
  4. D. HALES, R. L. WILLIAMSON, S. R. NOVASCONE, G. PASTORE, B. W. SPENCER, D. S. STAFFORD, K. A. GAMBLE, D. M. PEREZ, W. LUI. “BISON Theory Manual the Equations Behind Nuclear Fuel Analysis”,(INL/EXT-13-29930):4, Idaho National Laboratory, (2016).
  5. DENG, J. WEI, Y. WANG, and B. ZHANG, “Validation of the Fuel Rod Performance Analysis Code FRIPAC”, Nuclear Engineering and Technology, 51(6):1596–1609 (2019).
  6. J. FASSINO, A. ERICKSON, “REX: An Analytical Tool for Reactor Operating Envelope Expansion Through Fuel-Clad Thermo-Mechanics”, unpublished.
  7. LEPPäNEN, M. PUSA, T. VIITANEN, V. VALTAVIRTA, and T. KALTIAISENAHO. “The Serpent Monte Carlo Code: Status, Development and Applications in 2013”. Annals of Nuclear Energy, 82:142– 150, (2015).