Mathematical relation predicts achievable densities of compacted particles

Series of mathematical relationships predicts compact densities of spherical shapes in a cylinder as a function of particle dimension, and compact density of angular shapes as a function of particle shape and absolute size

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments

Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials

The binding epitopes of neurotrophin-3 to its receptors trkC and gp75 and the design of a multifunctional human neurotrophin.

Survival and maintenance of vertebrate neurons are influenced by neurotrophic factors which mediate their signal by binding to specific cell surface receptors. We determined the binding sites of human neurotrophin-3 (NT-3) to its receptors trkC and gp75 by mutational analysis and compared them to the analogous interactions of nerve growth factor (NGF) with trkA and gp75. The trkC binding site extends around the central beta-strand bundle and in contrast to NGF does not make use of non-conserved loops and the six N-terminal residues. The gp75 epitope is dominated by loop residues and the C-terminus of NT-3. A novel rapid biological screening procedure allowed the identification of NT-3 mutants that are able to signal efficiently through the non-preferred receptors trkA and trkB, which are specific for NGF and BDNF respectively. Mutation of only seven residues in NT-3 resulted in a human neurotrophin variant which bound to all receptors of the trk family with high affinity and efficiently supported the survival of NGF-, BDNF- and NT-3-dependent neurons. Our results suggest that the specificity among neurotrophic factors is not solely encoded in sequence diversity, but rather in the way each neurotrophin interacts with its preferred receptor

Technical basis for normal water chemistry guidelines: Review of laboratory studies of water chemistry effects on SCC (stress corrosion cracking)

The influence of dissolved oxygen, hydrogen, and various impurity anions on the stress corrosion cracking (SCC) susceptibility of sensitized Type 304 stainless steel (SS) and alternative piping materials such as Types 316NG and 347NG SS is being investigated in constant-extension-rate-tensile (CERT) tests and in cyclic loading experiments on fatigue precracked fracture-mechanics-type specimens at 289/sup 0/C. In these experiments, the crack growth behavior of the materials is being correlated with the impurity concentration and the electrochemical potentials of Type 304 SS and platinum electrodes in simulated BWR normal operating environments (approx. 200 to 300 ppb oxygen and < 100 ppb of various oxyanions of halides added as acid or salts, at a total conductivity of <1 ..mu..S/cm)

Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

Environmentally Assisted Cracking in Light Water Reactors. Semiannual Report, July 1998-December 1998.

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E &gt; 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments