FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes` integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6).

/pdf/frapcon-3-a-computer-code-for-the-calculation-of-steady-3op2iollzt.pdf

FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup

In-reactor deformation of cold-worked Zr–2.5Nb pressure tubes

Factors affecting the anisotropy of irradiation creep and growth of zirconium alloys

Irradiation creep and growth of zirconium alloys: A critical review

Cumulative zirconium and zirconium alloy creep data over a broad range of stresses (0.1 to 115 MPa) and temperatures (300 °C to 850 °C) were analyzed based on an extensive literature review and experiments. Zirconium obeys traditional power-law creep with a stress exponent of approximately 6.4 over stain rates and temperatures usually associated with the conventional “five-power-law” regime. The measured activation energies for creep correlated with the activation energies for zirconium self-diffusion. Thus, dislocation climb, rather than the often assumed glide mechanism, appears to be rate controlling. The common zirconium alloys (i. e., Zircaloys) have higher creep strength than zirconium. The stress exponents of the creep data in the five-power-law regime were determined to be 4.8 and 5.0 for Zircaloy-2 and Zircaloy-4, respectively. The creep strength of irradiated Zircaloy appears to increase relative to unirradiated material. It was found that the creep behavior of zirconium was not sensitive to oxygen in the environment over the temperature range examined.

P.A. Ross-Ross

Papers

The in-reactor creep of cold-worked Zircaloy-2 and Zirconium-2.5 wt % niobium pressure tubes

The relation between stress-relaxation and creep for some zirconium alloys during neutron irradiation