The first application of nonempirical molecular electronic structure theory to a realistic transition metal carbene complex is reported. The system chosen was (CO){sub 3}NiCH{sub 2}, methylene (tricarbonyl) nickel(0). All studies were carried out at the self-consistent-field (SCF) level. A large and flexibly contracted basis set was chosen, labeled Ni(15s 11p 6d/11s 8p 3d); C,O(9s 5p/4s 2p); H(5s/3s). In addition, the importance of methylene carbon d functions was investigated. The critical predicted equilibrium geometrical parameters were R [Ni-C (methylene)]=1.83 {Angstrom}, {theta}(HCH)=108°. The sixfold barrier to rotation about the Ni-C (methylene) axis is small, ~o.2 kcal. The electronic structure of (CO){sub 3}NiCH{sub 2} is discussed and compared with those of the "naked" complex NiCH{sub 2} and the stable Ni(CO){sub 4} molecule.

The equilibrium geometries of (CH{sub 3}){sub 2}Si=CH{sub 2} and H{sub 2}Si=CH{sub 2} have been determined at the self-consistent-field level of electronic structure theory using a double zeta basis set augmented with d functions on all heavy atoms. For the parent silaethylene, large scale configuration interaction (6920 configurations) demonstrates that electron correlation effects do not qualitatively alter the predicted structure. On this basis it is concluded that the experimental electron diffraction geometry of Mahaffy, Gutowsky, and Montgomery is likely to be seriously incorrect. Specifically the theoretical prediction for the dimethylsilaethylene Si=C distance is 1.692 {Angstrom}, while the range of experimental values presented was 1.815 - 1.835 {Angstrom}.

INTRODUCTION TO THE PROCEEDINGS OF THE SIXTH GEOTHERMAL RESERVOIR ENGINEERING WORKSHOP, STANFORD GEOTHERMAL PROGRAM Henry J. Ramey, Jr., and Paul Kruger Co-Principal Investigators Ian G. Donaldson Program Manager Stanford Geothermal Program The Sixth Workshop on Geothermal Reservoir Engineering convened at Stanford University on December 16, 1980. As with previous Workshops the attendance was around 100 with a significant participation from countries other than the United States (18 attendees from 6 countries). In addition, there were a number of papers from foreign contributors not able to attend. Because of the success of all the earlier workshops there was only one format change, a new scheduling of Tuesday to Thursday rather than the earlier Wednesday through Friday. This change was in general considered for the better and will be retained for the Seventh Workshop. Papers were presented on two and a half of the three days, the panel session, this year on the numerical modeling intercomparison study sponsored by the Department of Energy, being held on the second afternoon. This panel discussion is described in a separate Stanford Geothermal Program Report (SGP-TR42). This year there was a shift in subject of the papers. There was a reduction in the number of papers …

This article discusses the specificity of plant defences.

Article on the modulation of L-phenylalanine ammonia-lyase by pathway intermediates in cell suspension cultures of dwarf French bean (Phaseolus vulgaris L.).

The Oklo natural reactor was discovered in 1972. Experimental evidence has indicated that the fuel source was primarily 235U with a small contribution of 7 to 9 percent from the fission of 239Pu. This article's reevaluation of data indicates that 239Pu was an important source of fuel in some areas of the reactor. A small portion of xenon and krypton released from Oklo sample 1348 appears to have originated from a source enriched in 239Pu. That fuel source may have been the core of a natural breeder reactor.

A mathematical model of hydrogen fluoride (HF) deposition and accumulation of fluoride in a Eucalyptus rostrata forest has been developed. The model is based on tree physiology and meteorological principles. The data base for the model was derived from a literature survey of the physiological characteristics of E. rostrata and similar eucalyptus species and from current knowledge of meteorological processes in plant canopies.

The accumulation in groundwater of products from the radioactive decay of elements naturally found in rocks offers a potential for measuring the time that the groundwater has been contact with the rock. This method of dating groundwater has an advantage over using decay products from the atmosphere in that the amount of decay product becomes greater with increasing age rather than less. However, different decay products accumulate at different rates and, thus, have a different potential usefulness in age determinations. The most useful decay product is helium, produced from uranium and thorium. Argon produced from potassium is marginally useful for very old water. Neon, xenon and krypton are probably not useful because they are produced in extremely small quantities. In general, the potential for error increases when a long time to produce a small quantity is required.

Permeability tests conducted in the Climax stock granitic rock mass indicate that the bulk rock permeability can be highly variable. If moderately to highly fractured zones are encountered, the permeability values may lie in the range of 10{sup -4} to 10{sup -1} darcies. If, on the other hand, only intact rock or healed fractures are encountered, the permeability is found to be less than 10{sup -9} darcies. In order to assess the thermomechanical effect on fracture permeability, discrete fractures will be packed off and tested periodically throughout the thermal cycle caused by the emplacement of spent nuclear fuel in the Climax stock.

The status of LLNL Projects dealing with nuclear waste disposal in granitic rocks is reviewed. This review covers work done subsequent to the June 1979 Workshop on Thermomechanical Modeling for a Hardrock Waste Repository and is prepared for the July 1980 Workshop on Thermomechanical-Hydrochemical Modeling for a Hardrock Waste Repository. Topics reviewed include laboratory determination of thermal, mechanical, and transport properties of rocks at conditions simulating a deep geologic repository, and field testing at the Climax granitic stock at the USDOE Nevada Test Site.

In 1977 the Spent Fuel Handling and Packaging Program (SFHPP) was initiated by the Department of Energy to develop and test the capability to satisfactorily encapsulate typical spent fuel assemblies from commercial light-water nuclear power plants and to establish the suitability of one or more surface and near surface concepts for the interim dry storage of the encapsulated spent fuel assemblies. The E-MAD Facility at the Nevada Test Site, which is operated for the Department of Energy by the Advanced Energy Systems Division (AESD) of the Westinghouse Electric Corporation, was chosen as the location for this demonstration because of its extensive existing capabilities for handling highly radioactive components and because of the desirable site characteristics for the proposed storage concepts. This paper describes the remote operations related to the process steps of handling, encapsulating and subsequent dry storage of spent fuel in support of the Demonstration Program.

A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. …

Batch measurements provide an understanding of which experimental variables are important. For example, sorption ratios vary little with particle size (and surface area); however, groundwater composition and rock composition are quite important. A general correlation has been identified between mineralogy (major phases) and degree of sorption for strontium, cesium, and barium. Although these are approximate, a more detailed analysis may be possible as more samples are studied and the data base increased. Data from crushed tuff columns indicate that, except in simple cases where sorption coefficients are relatively low, and ion-exchange equilibria not only exist but are the dominant mechanism for removal of radioisotopes from solution, the simple relation between the sorption ratio R/sub d/ (or K/sub d/) and the relative velocity of radionuclides with respect to groundwater velocity may be insufficient to permit accurate modeling of the retardation of radionuclides. Additional work on whole core columns and larger blocks of intact material is required to better understand radionuclide sorption and transport through rock.

The Lawrence Livermore Laboratory is conducting a test of the geologic storage of encapsulated spent commercial reactor fuel assemblies in a granitic rock at the Nevada Test Site. The test, known as the Spent Fuel Test-Climax (SFT-C), is sponsored by the US Department of Energy, Nevada Operations Office. Eleven pressurized-water-reactor spent fuel assemblies are stored retrievably for three to five years in a linear array in the Climax stock at a depth of 420 m.

A test of retrievable dry geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.3 years out of reactor core (about 2 kW/canister thermal output) will be emplaced in a storage drift along with 6 electrical simulator canisters and their effects will be compared. Two adjacent drifts will contain electrical heaters, which will be operated to simulate within the test array the thermal field of a large repository. The test objectives, technical concepts and rationale, and details of the test are stated and discussed.

The Spent Fuel Test-Climax (SFT-C) is a test of dry geologic storage of spent nuclear reactor fuel. The SFT-C is located at a depth of 420 m in the Climax granitic stock at the Nevada Test Site. Eleven canisters of spent commercial PWR fuel assemblies are to be stored for 3 to 5 years. Additional heat is supplied by electrical heaters, and more than 800 channels of technical information are being recorded. The measurements include rock temperature, rock displacement and stress, joint motion, and monitoring of the ventilation air volume, temperature, and dewpoint.

Article discussing a multivariate analysis of health care services utilization among older Texans.

Paper that examines Tarman's investigation into gifted children in California.

Paper about the history of early admission programs in the United States of America. It proposes that advanced students not be kept to an age-in-grade pace, and that early admission programs be implemented to give these students a chance to learn at their own pace.

This note extends a certain combinatorics problem of I. Tomescu.

Article commenting on the prediction of gas chromatographic retention behavior with mixed liquid phases.

This article discusses electron capture from K shells by fully stripped ions.