Sample input and output data from a linear elastic finite element computer program (NOZ-FLAW).

From Abstract. "This report describes a linear elastic finite element computer program (NOZ-FLAW) which is used for the direct evaluation of mixed-mode stress intensity factors (K1, KII K111) along user-defined crack flaws at pressure-vessel nozzle corners." From Forward: "The work reported here was performed at the Oak Ridge National Laboratory and at the Georgia Institute of Technology under UCC-ND Subcontract No. 7565."

"This report describes a systematic study to determine the relationship between seismicity and geologic or geophysical features for the central and eastern United States in order to delineate seismic hazard."

FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

This report examines the adequacy of existing fire-hazards analysis methodology in the context of nuclear power plant safety

A report discussing work performed at the Oak Ridge National Laboratory in support of the ORNL Design Criteria for Piping and Nozzle Program being conducted for the U.S. Nuclear Regulatory Research.

The following data presents stress value results from finite-element mesh studies described in its parent document, Stresses in Reinforced Nozzle-Cylinder Attachments Under External Moment Loadings Analyzed by the Finite-Element Method: A Parameter Study.

This report presents recorded test data for Tests LLR-S0, LLR-3, LLR-4A, and LLR-5 of the thermal Fuels Behavior Program PBF/LOFT Lead Rod Test Series.

Power calibration and preconditioning data plots for PBF/LOFT Lead Rod Test Series, indicating the time after start of power calibration in hours with each variable.

Charts analyzing Multirod Burst Test (MRBT) Bundle B-5 test data produced at Oak Ridge National Laboratory.

Graphs presenting a summary of data for a fracture-mechanics parametric-type study conducted for a number of postulated transients, including critical fluences, pressures, and crack depth.

Report presenting the pretest physical characteristics and instrumentation of the Test PCM-1 (Power-Cooling Mismatch Test Series) fuel rod and shroud components.

Appendix to report presenting detailed data from the postirradiation examination and pretest characterization for Test PCM-1, including a description of the axial power profile.

Appendix to report presenting detailed data from the postirradiation examination and pretest characterization for Test PCM-1, including additional metallography illustrating the physical and chemical changes in the cladding and fuel.

Appendix to report presenting detailed data from the postirradiation examination and pretest characterization for Test PCM-1, including tabulated data from the particle size analysis of the broke fuel rod pieces.

Appendix to report presenting detailed data from the postirradiation examination and pretest characterization for Test PCM-1, including details of the microprobe examination.

Appendix to report presenting detailed data from the postirradiation examination and pretest characterization for Test PCM-1, including changes in cladding thermal conductivity with oxidation to ZrO₂.

Appendix to report presenting detailed data from the postirradiation examination and pretest characterization for Test PCM-1, including the U0₂ oxidation calculation.

This report presents the analysis of the fuel rod materials behavior based on the postirradiation examination of the Test PCM-1 fuel rod from the Power-Cooling-Mismatch Test Series.

Plots of qualified data from Reactivity Initiated Accident (RIA) Series 1-4 fuel behavior test conducted at the Idaho National Engineering Laboratory in Idaho Falls, Idaho.

Report summarizing the design of Power Burst Facility (PBF) test rods, test train hardware, the individual rod assembly instrumentation, and test conduct.

Report presenting the pretest fuel rod characterization data for four rods used in Test RIA 1-2.

Report providing a detailed description of the Power Burst Facility (PBF) design and capabilities, including an illustration of the axial cross section of the PBF in-pile tube.

Report describing four different methods of measuring the total radially averaged fission energy deposited in PBF (Power Burst Facility) fuel rods, as well as a summary evaluation of each method.