Call Number (LC) Title Results
E 1.99:ornl/nureg-11 Decay of ⁹⁹Mo 1
E 1.99:ornl/nureg-13 Kinetic model for predicting the composition of chlorinated water discharged from power plant cooling systems 1
E 1.99:ornl/nureg-14 Fission-product energy release for times following thermal-neutron fission of ²³⁵U between 2 and 14000 seconds 1
E 1.99:ornl/nureg-15 Significance of reheat cracks to the integrity of pressure vessels for light-water reactors 1
E 1.99:ornl/nureg-16 Model of iodine transport and reaction kinetics in a nuclear fuel reprocessing plant 1
E 1.99:ornl/nureg-17 Zirconium metal-water oxidation kinetics. IV. Reaction rate studies. [BWR PWR] 1
E 1.99:ornl/nureg-18/v1 Stress analysis of cylindrical pressure vessels with closely spaced nozzles by the finite-element method. Volume 1. Stress analysis of vessels with two closely spaced nozzles under internal pressure. [BWR; PWR; MULT-NOZZLE code] 1
E 1.99: ornl/nureg-18/v2 Stress analysis of cylindrical pressure vessels with closely spaced nozzles by the finite-element method. Volume II. Vessels with two nozzles under external force and moment loadings. [MUFT-NOZZLE code] 1
E 1.99:ornl/nureg-19 PWR blowdown heat transfer separate-effects program data evaluation report system response for thermal-hydraulic test facility test series 100. 1
E 1.99:ornl/nureg-22 Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis. 1
E 1.99:ornl/nureg-23 ORINC a one-dimensional implicit approach to the inverse heat conduction problem. [PWR] 1
E 1.99:ornl/nureg-24 Experimental study of plastic responses of pipe elbows 1
E 1.99:ornl/nureg-31 Zirconium metal-water oxidation kinetics. V. Oxidation of Zircaloy in high pressure steam. [PWR] 1
E 1.99:ornl/nureg-32 Validation of the kinetic model for predicting the composition of chlorinated water discharged from power plant cooling systems 1
E 1.99:ornl/nureg-35 Performance testing of single electrically heated fuel pin simulators for PWR LOCA experiments 1
E 1.99:ornl/nureg-37 Corrosion of steel tendons in concrete pressure vessels review of recent literature and experimental investigations. 1
E 1.99: ornl/nureg-39 Delayed beta- and gamma-ray production due to thermal-neutron fission of /sup 235/U, spectral distributions for times after fission between 2 and 14,000 sec tabular and graphical data. 1
E 1.99:ornl/nureg--42 The Use of Evaporation to Treat Radioactive Liquids in Light-Water-Cooled Nuclear Reactor Power Plants 1
E 1.99: ornl/nureg-60 Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding significance for risk assessment. 1
E 1.99: ornl/nureg-70 Radionuclide decay data base - index and summary table 1