A stress analysis of the piping modification, resulting from relocation of heat exchanger BS-H-31 of the LOFT Blowdown Suppressing Tank Spray System, was performed. The piping, fittings, and supports were found to comply with the criteria of Section III of the ASME Boiler and Pressure Vessel Code, 1974.

Detailed consideration of several aspects of a field-reversal experiment was begun in the Mirror Fusion Test Facility (MFTF): Model calculations have provided some plausible parameters for a field-reversed deuterium plasma in the MFTF, and a buildup calculation indicates that the MFTF neutral-beam system is marginally sufficient to achieve field reversal by neutral injection alone. However, the many uncertainties indicate the need for further research and development on alternate buildup methods. A discussion of experimental objectives is presented and important diagnostics are listed. The range of parameter space accessible with the MFTF magnet design is explored, and we find that with proper aiming of the neutral beams, meaningful experiments can be performed to advance toward these objectives. Finally, it is pointed out that if we achieve enhanced n tau confinement by means of field reversal, then quasi-steady-state operation of MFTF is conceivable.

The LOFT pressurizer spray and surge line piping was analyzed for a 450/sup 0/F step change in fluid temperature. This transient was chosen to conservatively represent several pressurizer operating transients that had not previously been analyzed. These include temperature transients resulting from a 300/sup 0/F ..delta..T between pressurizer temperature and cold leg temperature, injection of a cooled (70/sup 0/F) slug of stagnant fluid into the hot (540/sup 0/F) spray line piping, and inflow of 100/sup 0/F primary coolant system water into the hot (480/sup 0/F) surge line piping.

The purpose of this test was to demonstrate that the System Pump utilized for the developmental program to be conducted on the Kilowatt Isotope Power System (KIPS) fulfilled the requirements of Test Procedure 398, Component Test Procedure for the GDS System Pump. Results of the testing performed on the System Pump are presented.

Laboratory measurements to determine the sorption distribution coefficients, Kd, of radionuclides present in, and potentially leached from, radioactive wastes, in contact with representative geologic media, have been conducted. The nuclides studied include Cs, Sr, Tc, Ru, Sb, Ce, Eu, Pu, Np, Cm, Am, U, and Pa. The crushed rock materials used were from the vicinity of the Waste Isolation Pilot Plant, a project to isolate radioactive wastes in a bedded salt facility, near Carlsbad, New Mexico. Solutions used consist of salt brine and groundwater, specific to the WIPP site, plus distilled water, for laboratory intercomparisons. The batch Kd data reported, plus data from sorption and migration measurements being conducted or planned elsewhere, will be used to evaluate the potential for radionuclide migration from the bedded salt WIPP facility. The data can be used for transport modeling and for safety assessment determinations.

The multiblade slurry technique capable of slicing 10 cm ingot into wafers 0.25 mm thick with only 0.20 mm kerf loss and 98% yield has been demonstrated. The total silicon requirement represents an ingot to sheet conversion of 0.95 m/sup 2//kg. Full production slicing tests have demonstrated the cost of MS slicing to contribute $40 to $50/m/sup 2/, with ''best effort'' estimates for today's configuration to be $30 to $35/m/sup 2/. By reducing material cost, and increasing the specific capacity of a saw to slice 900 wafers simultaneously, the long-term cost of MS slicing is estimated to be less than $10/m/sup 2/. The conversion of ingot to sheet is shown to be the most valuable contribution of slicing technology. At today's ingot costs, and with the thin wafer, low kerf loss slicing techniques demonstrated, the silicon material represents 5 to 10 times the cost of the wafering process in finished silicon wafers. Increasing the number of blades used in MS slicing from 100 to 150 to 225 to 300 has resulted in a reduction of yield to 33 to 70% for thin slicing, or an increase in wafer thickness to 0.30 mm slices. The limitation is intrinsic misalignment of multiple …

The program documentation is presented for the FCHART computer model. This document highlights and explains the various components comprising the FCHART interactive program. Specifically, this document includes descriptions of: program narrative, program overview, model components, thermal analysis, and economic analysis. FCHART is a computer model which calculates domestic water and space heating loads for residential and commercial buildings. This thermal analysis of performance standards of active solar space and domestic water heating systems can be accomplished for either water or air as the transfer fluid. FCHART is also capable of performing life-cycle cost analyses of solar space and water heating of various levels of intricacy. FCHART is written in FORTRAN II. Functions are as follows: CALC and ECON.

In addition to data for the first three quarters of 1977, cumulative data since plant startup are presented, along with a listing of decayed activity to the various plant burial sites. (DLC)

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The new algorithm used to calculate the ponderomotive force in LASNEX is discussed. The algorithm is shown to be identical with the corresponding analytic expression when the WKB solution for the fields in a plane-parallel medium is used. LASNEX calculations performed to check the ponderomotive force algorithm are described. 2 figures.

The I/O instructions for the Intel 8008 Microprocessor have versatilities that are not apparent to most users. The Standard lament is that there are only 8/sub 10/ input instructions. The fact that there are only 24/sub 10/ output instructions is sometimes belabored, but 24 is normally considered sufficient for most microprocessor applications. This note discusses ways to expand the available number of both these instructions. 4 figures.

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The seeded growth furnace with a slotted molybdenum susceptor has been operated during the past quarter. The susceptor design permits flat sheets of refractory materials to be used in making up the tray. Fused quartz and sapphire trays were used. Local surface cooling was obtained by means of a water-cooled cold finger. Seeded surface growth has been obtained in this furnace using a seed with (111) axis and (112) principal faces. The silicon grown out of solution has a region approximately 3 mm x 5 mm, which is substantially single-crystal with the same crystallographic orientation as the seed. Sessile drop experiments have been initiated as a rapid method of determining the suitability of various refractory materials as trays for tin melts. These experiments indicate the presence of some impurity in the starting tin, which forms a surface film that prevents the tin from wetting silicon. This film can be removed by heating the tin to 900/sup 0/C in hydrogen.

This report provides necessary user information to implement and use a graphics package for the CORTES finite-element computer programs. Complete input instructions are provided. Sample input and output are given.

The performance of a gamma-ray densitometer system using computer modeling techniques is predicted. The system studied consists of a /sup 60/Co calibration source, a pipe containing a variable amount of water absorber where the water contains radioactive /sup 16/N, and a shielded and collimated NaI detector system. The basic question is: Can one detect the change in signal from the collimated /sup 60/Co calibration source caused by a water density change when the detector system is subject to an adjacent background source equivalent to about 30 Ci of high energy (avg. 6 MeV, pk. 10 MeV) gamma rays. The first report indicated that excessive (saturation) detector count rates could easily occur from this high energy, high intensity background and the second basic question is: How much lead shielding is required for typical NaI detectors to keep the detector count rate from this source below 50 kcps (50,000 counts per second). The importance of this question is borne out by the results that indicate that the weight of lead shielding required is about 1000 lbs.

One finds the fourth-order differential equation describing an Alfven-ion-cyclotron wave propagating along a magnetic field of varying intensity. The equation is self-adjoint and possesses non-trivial turning points. The final form of the equation is checked using MACSYMA, a system for performing algebra on a computer.

Continuing studies of fission cross-section ratios at Lawrence Livermore Laboratory have led to the measurement of the /sup 243/Am//sup 235/U fission cross-section ratio from 0.1 to 30 MeV. A value of 1.429 +- 0.037 was obtained for the average cross-section ratio from 1.75 to 4.00 MeV by use of the threshold method. 2 figures, 1 table.

Rf bypass capacitors were added to the Head Position Preamplifier and Servo Power Amplifier on Digital Equipment Corporation (DEC) RK05 DISK TRANSPORTS, and ground straps were installed between all peripheral chassis and the main frame of the minicomputers to eliminate interference that was generated by hand-held two-way radios operating in the 165 and 410 MHz ranges. 2 figures.

This report contains 679 abstracts from the Nuclear Safety Information Center (NSIC) computer file dated 1973 through 1976 covering material properties with respect to structural integrity. All materials important to the nuclear industry (except concrete) are covered for mechanical properties, chemical properties, corrosion, fracture or failure, radiation damage, creep, cracking, and swelling. Keyword, author, and permuted-title indexes are included for the convenience of the user.

CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP.

The purpose of this research program is the demonstration of the technical feasibility of a high-temperature, solid-electrolyte fuel cell, building on the presently-existing technological base. The bulk of the work consists of identifying and selecting candidate component materials, fabricating cells and cell stacks from these materials, and testing these cells and cell stacks under conditions which might be expected in operation. The major emphasis is on the invention and fabrication of an interconnection film, which has the following properties at 1000/sup 0/C, the proposed temperature of stack operation: (1) chemical stability in air and fuel gas; (2) electronic resistivity in the working environment of less than 50 ohm-cm; (3) thermal expansion characteristics compatible with other components; and (4) non-reactivity with adjoining components. Emphasis is also being placed on the fabrication of porous support tubes of calcia-stabilized zirconia, to the desired specifications of tensile strength, porosity, surface finish, and purity. Concurrently, tubes will be prepared in-house and also will be purchased from commercial suppliers and evaluated for compliance with the following specifications: (1) tensile strength of 5000 to 10,000 psi; (2) open porosity of 25 to 30 volume percent; and (3) no surface pore greater than 10 ..mu..m in diameter. Progress …

The U.S. Atomic Energy Commission has defined transuranic-contaminated waste as material that is contaminated with greater than 10 nCi/g of long-lived and highly toxic radionuclides. The analyses and techniques that show potential for detection at this low level are reviewed. The physics involved with the methods described is relatively straightforward, but difficulties arise in the actual application of the techniques. Therefore, discussion is concentrated on the application of the analyses, including the necessary calculational considerations and the potential problems that may be encountered. No simple method presently exists for measuring transuranic isotopes in solid waste material at the low level of 10 nCi/g. In the absence of an interfering background (< 1 mR/hr), gamma-ray spectroscopy is the best method available. Monitoring helium production in sealed waste material shows good potential for detection at the 10-nCi/g level. The only other viable method involves a complex procedure of counting spontaneous-fission neutrons, neutron irradiation, and counting delayed neutrons. However, if the minimum detectable level permitted in waste material were raised to ..mu..Ci/g, we would be able to use several more measurement methods detect almost all the transuranic isotopes present in solid waste.

Preliminary experiments studying the feasibility of gamma-ray spectrometry for in-line measurements of plutonium in solutions are described. The isotopic content of the plutonium was measured by direct counting and the total plutonium content by a differential attenuation technique. Two separate experiments on different solutions were performed with each technique. Our results show that both isotopic and total measurements can be made with precisions around 0.25 to 0.5 percent.

The Office of Waste Isolation program was reviewed. Topics involving program management and geologic aspects of radioactive waste isolation were discussed.