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QUARTERLY STATUS REPORT ON LAMPRE PROGRAM FOR PERIOD ENDING FEBRUARY 20, 1961

Description: S>The LAMPRE-I project is summarized in terms of capsule development and production, sodium system, cover gas system, capsule charge, shielding, and fuel storage facility. The loading of the LAMPRE-I core was begun on January 20, 1961 with the sodium temperature set at 160 deg C. The reactor was brought to criticality on February 17, 1961. Operation of the Sodium Test Facility was continuous ex cept for 6 maintenance and inspection shutdowns resulting in 680 idle hours. The intermediate sodium heat exchanger, steam generating unit, centrifugal sodium pumps, sodium flow control valves, and gas-fired sodium heater are discussed. Heat transfer test results are given for the various components. Research and development activities for the LAMPRE program are reported in the topics fuel and alloy program, container alloy development, direct contact core studies, development of liquid fuels, container materials for reactor fuels, and fuel reprocessing. (M.C.G.)
Date: March 1, 1961
Partner: UNT Libraries Government Documents Department

QUARTERLY STATUS REPORT ON LAMPRE PROGRAM FOR PERIOD ENDING MAY 20, 1961

Description: LAMPRE-L The reactor was started up and held st 160 deg C and at hot critical for various measurements. Some of the 160 deg C measurements included rod worth, temperature coefficient, and shim-down extrapolated mass. In the approach to hot critical, multiplication was measured during the temperature increase from 160 to 500 deg C and found to increase markedly when the core meltsd. Reactivity changes due to jarring of the fuel capsules to remove bubbles and voids in the melted fuel were observed. Zero measurements made at 48O deg C are reported. Power coefficient measurements at 0 to 46 kw and power demand experiments were made, and the 50-kw transfer function is plotted. Shielding inadequacies are discussed briefly. The results of a fuel pin scan are given. Observations on the performance of the sodium system are discuaaed. Type 17-4 PH stainless steel was found to undergo additional grain boundary precipitution during long-term exposure to temperatures below the hardening temperature, and measurements of its properties are reported. fuel and Container Development. Impurities in the LAaMPRE-1 fuel which has a corrosive effect on the fuel container are discussed. Observations in the Pu -Ce-Co phase diagram are reported. Different types of LAMPRE-1 capsule failure are discussed. Work on improving the corrosion resiatance of Ta revealed Y and W to be beneficial additives. Sodium Test Facility. The test facility was opersted nearly continuously. with test equipment pieces receiving total operating times as high as 10,400 hr. The He cover gas in the primary and secondary loops was analyzed for buildup of O/sub 2/, H/sub 2/, and CH/sub 4/. (D.L.C.)
Date: June 1, 1961
Partner: UNT Libraries Government Documents Department

QUARTERLY STATUS REPORT ON LAMPRE PROGRAM FOR PERIOD ENDING NOVEMBER 20, 1960

Description: A decision was made to begin fabrication for the initial core loading of LAMPRE-1 with capsules from the tantalum on hand. Fuel for the first loading will be the cast Fe--Pu alloy from LCX III capsules and will contain carbon and stabilizer. Certification and melt-freeze tests are continuing on LAMPRE type capsules. The filling of the reactor sodium system is described. The cover gas system operated satisfactorily during the sodium shakedown phase. Four of the 15 core thermocouples have operated improperly since the sodium system was filled. The capsule charges were operated to remove dummy capsules and insert tantalum capsules containing test coupons. The 2000-kw Sodium Test Facility, including test steam generator, was operated continuously from Aug. 20 to Nov. 20, except for l59 hr of shutdown required for maintenance of auxiliary steam system equipment. Mercury-water flow systems were set up and are being operated to study both lift and jet pumping. A second fuel pumping experiment using Co --Ce - -Pu fuel was set up and tried without success. The effects of various additives on the properties of Fe-Pu fuels are being studied. The fabrication of LAMPRE-1 capsules by impact-extruding a rod-slug into a starting cup followed by six ironing stages is described. Materials that were corrosion tested as fabricated capsules include arcmelted and electron-beann-melted high-purity tantalum and Ta-- 0.1 wt.% W alloy. Corrosion tests are in progress on experimental deep-drawn capsules made from Ta --0.1 wt. % W--0.2 wt. % Y. An x-ray fluorescence spectrographic method was developed for determining hafnium in Ta--Hf and Ta--W- Hf alloys. Work is in progress on the development of a solvent extraction method for the recovery of plutonium residues from various pyrometallurgical processes. (For preceding period see LAMS-2462.) (W.L.H.)
Date: December 1, 1960
Partner: UNT Libraries Government Documents Department

QUARTERLY STATUS REPORT ON LAMPRE PROGRAM FOR PERIOD ENDING NOVEMBER 20, 1963

Description: The development and operation of the Los Alamos Molten Plutonium Reactor Experiments are described. The development and compatibility of iron-- plutonium and cerium-cobalt-- plutonium alloy fuels are evaluated. The fabrication and testing of the liquid sodiam loop are summarized. The phase studies of plutoniam alloys containing neodymium, scandium, yttrium, praseodymium, and cerium-- cobalt mixtures are reported. (N.W.R.)
Date: December 1, 1963
Partner: UNT Libraries Government Documents Department

Transuranic Solid Waste Management Programs. Progress report, July-- December 1974

Description: Progress is reported for three transuranic solid waste management programs funded at the Los Alamos Scientific Laboratory by the Energy Research and Development Administration Division of Waste Management and Transportation. Under the Transuranic Waste Research and Development Program, a completed evaluation of stainless steel drums showed that although the material has superior corrosion-resistant properties, its higher cost makes a thorough investigation of other container systems mandatory. A program to investigate more economical, nonmetallic containers is proposed. Preliminary fire tests in mild steel drums have been completed with fire propagation not appearing to be a problem unless container integrity is lost. Investigation of the corrosion of mild steel drums and the evaluation of potential corrosion inhibitors, in a variety of humid environments, continues. Experimental results of both laboratory and field investigations on radiolysis of transuranic elements in hydrogenous waste are discussed. Progress in the development of instrumentation for monitoring and segregating low-level wastes is described. New plans and developments for the Transuranic-Contaminated Solid Waste Treatment Development Facility are presented. The current focus is on a comparison of all alternative waste reduction systems toward a relative Figure of Merit with universal application. Drawings, flowsheets, and building layouts are included, and the proposed incinerator device is detailed. The release mechanisms, inter- and intraregional transport mechanisms, and exhumation studies relevant to the Evaluation of Transuranic-Contaminated Radioactive Waste Disposal Areas Program are defined and analyzed. A detailed description is given of the formulation of the computer simulation scheme for the intraregional biological transport model. (auth)
Date: October 1, 1975
Partner: UNT Libraries Government Documents Department

ENDF/B-IV fission-product files: summary of major nuclide data

Description: The major fission-product parameters [sigma/sub th/, RI, tau/sub 1/2/, E- bar/sub $beta$/, E-bar/sub $gamma$/, E-bar/sub $alpha$/, decay and (n,$gamma$) branching, Q, and AWR] abstracted from ENDF/B-IV files for 824 nuclides are summarized. These data are most often requested by users concerned with reactor design, reactor safety, dose, and other sundry studies. The few known file errors are corrected to date. Tabular data are listed by increasing mass number. (auth)
Date: September 1, 1975
Creator: England, T.R. & Schenter, R.E.
Partner: UNT Libraries Government Documents Department

Electron elastic scattering cross sections from 1 keV to 100 MeV for elements Z = 1 to 100

Description: Tables of electron elastic scattering differential cross sections of elements (Z = 1 to 100) are given for electron energies in the range of 1 keV to 100 MeV and scattering angles in the range of 1 to 179$sup 0$. The function describing the asymmetry of a polarized electron beam after the scattering process is also included in the tables. (auth)
Date: April 1, 1975
Creator: Storm, E. & Hancock, J.H.
Partner: UNT Libraries Government Documents Department

Experience related to the safety of advanced LMFBR fuel elements

Description: Experiments and experience relative to the safety of advanced fuel elements for the liquid metal fast breeder reactor are reviewed. The design and operating parameters and some of the unique features of advanced fuel elements are discussed breifly. Transient and steady state overpower operation and loss of sodium bond tests and experience are discussed in detail. Areas where information is lacking are also mentioned. (auth)
Date: July 1, 1975
Creator: Kerrisk, J.F.
Partner: UNT Libraries Government Documents Department

Application of sensitivity analysis to a quantitative assessment of neutron cross-section requirements for the TFTR: an interim report

Description: A computational method to determine cross-section requirements quantitatively is described and applied to the Tokamak Fusion Test Reactor (TFTR). In order to provide a rational basis for the priorities assigned to new cross- section measurements or evaluations, this method includes quantitative estimates of the uncertainty of currently available data, the sensitivity of important nuclear design parameters to selected cross sections, and the accuracy desired in predicting nuclear design parameters. Perturbation theory is used to combine estimated cross-section uncertainties with calculated sensitivities to determine the variance of any nuclear design parameter of interest. (auth)
Date: September 1, 1975
Creator: Gerstl, S.A.W.; Dudziak, D.J. & Muir, D.W.
Partner: UNT Libraries Government Documents Department

Automated spectrophotometer for plutonium and uranium determination

Description: The automated spectrophotometer described is the first in a planned series of automated instruments for determining plutonium and uranium in nuclear fuel cycle materials. It has a throughput rate of 5 min per sample and uses a highly specific method of analysis for these elements. The range of plutonium and uranium measured is 0.5 to 14 mg and 1 to 14 mg, respectively, in 0.5 ml or less of solution with an option to pre-evaporate larger volumes. The precision of the measurements is about 0.02 mg standard deviation over the range corresponding to about 2 rel percent at the 1-mg level and 0.2 rel percent at the 10-mg level. The method of analysis involves the extraction of tetrapropylammonium plutonyl and uranyl trinitrate complexes into 2-nitropropane and the measurement of the optical absorbances in the organic phase at unique peak wavelengths. Various aspects of the chemistry associated with the method are presented. The automated spectrophotometer features a turntable that rotates as many as 24 samples in tubes to a series of stations for the sequential chemical operations of reagent addition and phase mixing to effect extraction, and then to a station for the absorbance measurement. With this system, the complications of sample transfers and flow-through cells are avoided. The absorbance measurement system features highly stable interference filters and a microcomputer that controls the timing sequence and operation of the system components. Output is a paper tape printout of three numbers: a four-digit number proportional to the quantity of plutonium or uranium, a two-digit number that designates the position of the tube in the turntable, and a one-digit number that designates whether plutonium or uranium was determined. Details of the mechanical and electrical components of the instrument and of the hardware and software aspects of the computerized control system are provided.
Date: September 1, 1975
Creator: Jackson, D.D.; Hodgkins, D.J.; Hollen, R.M. & Rein, J.E.
Partner: UNT Libraries Government Documents Department

Beginning-of-life neutronic analysis of a 3000-MW(t) HTGR

Description: The results of a study of safety-related neutronic characteristics for the beginning-of-life core of a 3000-MW(t) High-Temperature Gas-Cooled Reactor are presented. Emphasis was placed on the temperature-dependent reactivity effects of fuel, moderator, control poisons, and fission products. Other neutronic characteristics studied were gross and local power distributions, neutron kinetics parameters, control rod and other material worths and worth distributions, and the reactivity worth of a selected hypothetical perturbation in the core configuration. The study was performed for the most part using discrete-ordinates transport theory codes and neutron cross sections that were interpolated from a four-parameter nine-group library supplied by the HTGR vendor. A few comparison calculations were also performed using nine-group data generated with an independent cross-section processing code system. Results from the study generally agree well with results reported by the HTGR vendor. (auth)
Date: December 1, 1975
Creator: Vigil, J.C.
Partner: UNT Libraries Government Documents Department

Chemistry computations for irradiated hot air

Description: A description is given of a computational model of chemical kinetics in air at temperatures between 300 and 4000$sup 0$K, with and without imposed fluxes of ionizing radiation and uv radiation, at pressures up to 10 atmospheres. Included are 1360 chemical reactions, involving 71 H, C, N, and O-containing chemical species. The reaction set is complete in the sense of including a reverse reaction for every reaction and including the dominant destruction reactions for each species produced. Photochemical reaction rates are computed in terms of a prescribed intensity and spectral distribution of radiant flux. Reactions of the metastable species O($sup 1$D), N($sup 2$D), O$sub 2$($sup 1$$delta$g), and O$sub 2$($sup 1$$Sigma$$sup +$/sub g/) are included explicitly, but all other quantum state populations are assumed to be in thermodynamic equilibrium. Results of several sets of model computations are described. These include: computations of rates of equilibration of systems subjected to abrupt temperature changes at a pressure of 1 atmosphere; computations of relaxation of systems subjected to impulsive sources of ionizing radiation at P = 1 atmosphere and 0.01 atm, with and without superimposed sources of optical-uv radiation; and computations of the approach to steady state in systems subjected to steady sources of ionizing radiation, for pressures between 10$sup -3$ atm and 10 atm. For computing concentrations of electrons, ions, NO, NO$sub 2$, O, N, and O$sub 3$, an abbreviated lumped-parameter reaction set was constructed with 31 reactions, designed to be incorporated in a fluid-dynamics code. Rate coefficients were calibrated against results from the 1360-reaction code for temperatures between 300 and 4000$sup 0$K and pressures between 10$sup -3$ and 10 atm. (auth)
Date: August 1, 1975
Creator: Sutherland, C.D. & Zinn, J.
Partner: UNT Libraries Government Documents Department

CHIT: a cost accounting program for postirradiation examinations of fast breeder reactor materials

Description: CHIT is the Los Alamos Scientific Laboratory's cost accounting computer program for nondestructive and destructive examinations of irradiated fuel pins. The program allows immediate retrieval of fuel pin examination information and provides itemized listings for completed and projected fuel pin examinations, detailed cost accounting summaries for each investigator, fuel pin examinations during a specified time interval, and various subsets of the information. CHIT has been in successful operation for the past two years, providing precise information on cost accounting more efficiently than possible with a manual technique. (auth)
Date: September 1, 1975
Creator: Phillips, J.R. & Dowler, K.E.
Partner: UNT Libraries Government Documents Department

Conceptual design for a fast reactor safety test facility. Preliminary report

Description: In May 1975 Los Alamos Scientific Laboratory issued a preliminary report on a study of Fast Reactor Safety Test Facilities. The study addressed itself to three closely related tasks. (1) A review of the current understanding of fast reactor safety with the aim of identifying important areas of uncertainty which cannot be adequately resolved using analysis, out of pile and/or existing in-pile facilities. (2) Conceptual design studies of one or more new in-pile facilities having characteristics identified in (1) above. (3) An examination of advanced data acquisition techniques for possible incorporation in the new facilities. The work reported is an extension of the earlier work in task area (2) above. Based largely on conclusions drawn from the earlier work the scope of the current effort has been narrowed to the design study of a Type A facility operating in the Class III mode, i.e., a facility capable of accommodating up to 37 test pins and capable of imposing a burst on top of a high steady state power level. (auth)
Date: August 1, 1975
Creator: Allen, J.D.; Cort, G.E.; McLaughlin, T.P. & Palmer, R.G.
Partner: UNT Libraries Government Documents Department

Consistency among differential nuclear data and integral observations: the ALVIN code for data adjustment, for sensitivity calculations, and for identification of inconsistent data

Description: Successful nuclear design requires adequate prediction of integral design parameters, and this in turn requires an adequate differential nuclear data base. Data bases that apparently permit reduced biases and design margins have been developed by a) least squares adjustment of differential data or b) trial-and-error selection from alternative evaluated data sets. Criticisms and defenses of such procedures are discussed. Useful data adjustment is related to consistency of the combined differential-integral data set and consistency tests related to least squares adjustment procedures are described. An approach to data adjustment is suggested that is contingent on consistency analysis. A FORTRAN code ALVIN has been developed to carry out the indicated data consistency and adjustment calculations, and to compute required sensitivities of integral parameters to nuclear data changes. The sensitivity modules of ALVIN are validated by computing with two distinct methods the cross-section sensitivity profile for neutron penetration through a thick iron shield. The data consistency and adjustment modules DAFT2 (for arbitrary variance-covariance data) and DAFT3 (for differential data base of arbitrary size uncorrelated with integral data) are validated by comparing their results for a set of data for three ZPR criticals. 2 figures, 5 tables, 51 references (auth)
Date: May 1, 1975
Creator: Harris, D.R.; Reupke, W.A. & Wilson, W.B.
Partner: UNT Libraries Government Documents Department

Interaction of $sup 238$PuO$sub 2$ heat sources with terrestrial and aquatic environments

Description: Radioisotope thermoelectric generators used in space missions are designed with a great factor of safety to ensure that they will withstand reentry from orbit and impact with the earth, and safely contain the nuclear fuel until it is recovered. Existing designs, utilizing $sup 238$PuO$sub 2$ fuel, have proved more than adequately safe. More data about the interaction of this material with terrestrial and aquatic environments is continually being sought to predict the behavior of these heat sources in the extremely unlikely contact of these materials with the land or ocean. Terrestrial environments are simulated with large environmental chambers that permit control of temperature, humidity, and rainfall using different kinds of soils. Rain falling on thermally hot chunks of $sup 238$PuO$sub 2$ causes the spallation of the surface of the fuel into extremely fine particles, as small as 50 nm, that are later transported downward through the soil. Some of the plutonia particles become agglomerated with soil particles. Plutonium transport is more significant during winter than during summer because evaporation losses from the soil are less in winter. Aquatic environments are simulated with large aquaria that provide temperature and aeration control. Earlier fuel designs that employed a plutonia-molybdenum cermet showed plutonium release rates of about 10 $mu$Ci/m$sup 2$ - s, referred to the total surface area of the cermet. Present advanced fuels, employing pure plutonium oxide, show release rates of about 20 nCi/m$sup 2$ - s in seawater and about 150 nCi/m$sup 2$ - s in freshwater. The temperature of these more advanced heat sources does not seem to affect the release rate in seawater. (auth)
Date: January 1, 1975
Creator: Patterson, J.H.; Nelson, G.B.; Matlack, G.M. & Waterbury, G.R.
Partner: UNT Libraries Government Documents Department

Investigation of pattern recognition techniques for the indentification of splitting surfaces in Monte Carlo particle transport calculations

Description: Statistical and deterministic pattern recognition systems are designed to classify the state space of a Monte Carlo transport problem into importance regions. The surfaces separating the regions can be used for particle splitting and Russian roulette in state space in order to reduce the variance of the Monte Carlo tally. Computer experiments are performed to evaluate the performance of the technique using one and two dimensional Monte Carlo problems. Additional experiments are performed to determine the sensitivity of the technique to various pattern recognition and Monte Carlo problem dependent parameters. A system for applying the technique to a general purpose Monte Carlo code is described. An estimate of the computer time required by the technique is made in order to determine its effectiveness as a variance reduction device. It is recommended that the technique be further investigated in a general purpose Monte Carlo code. (auth)
Date: August 1, 1975
Creator: Macdonald, J.L.
Partner: UNT Libraries Government Documents Department

Laser-fusion target fabrication: application of a polymeric ablator coating to a ball-and-disk target design by the physical vapor deposition of polyethylene

Description: A technique for applying polyethylene by physical vapor deposition is described. The ball-and-disk target design requires the application of a thin film of polyethylene on the front surface of the ball and substrate upon which the ball is mounted. Disk-shaped films, typically 20-$mu$m-diam by 1-$mu$m- thick, are successfully applied by this method. (auth)
Date: October 1, 1975
Creator: Simonsic, G.A.
Partner: UNT Libraries Government Documents Department

LINX and BINX: CCCC utility codes for the MINX multigroup processing code

Description: The LINX and BINX codes were written to manipulate multigroup cross- section libraries in the CCCC format. The LINX code is used to merge two ISOTXS or BRKOXS libraries. The BINX code is used to convert any ISOTXS, BRKOXS, or DLAYXS library from binary to BCD mode or back with the option to list all or any part of the library. These codes are utilities for the MINX multigroup processing system. 2 tables (auth)
Date: January 1, 1976
Creator: MacFarlane, R.E. & Kidman, R.B.
Partner: UNT Libraries Government Documents Department

Model for predicting the redistribution of particulate contaminants from soil surfaces

Description: A computerized model was developed to describe the redistribution of wind eroding soil-contaminant mixtures. Potentially mobile particulate contaminants can, in the first approximation, be assumed to be indistinquishable from the wind eroding soil in which they are distributed. A grid network characterizes important soil and surface conditions, and mass conserving control volumes are constructed on each cell. Material is transported through the vertical and top surfaces of a control volume by a modified Bagnold-Chepil horizontal flux formulation and modified Gillette vertical flux formulation, respectively. The vertical emissions, considered as puffs from area sources, create at regular time intervals a contaminant cloud which is proportional to the suspendable ground concentration. These puffs diffuse downwind under time- dependent wind velocity and atmospheric stability conditions, maintaining during the time interval a three-dimensional Gaussian distribution of concentration with cloud volume. Material from each puff is deposited in downward cells, leading to the possibility of many different flights from these new sources. The usefulness of this predictive tool is demonstrated by calculations involving mixtures of particulate $sup 238$PuO$sub 2$ in highly erodible soils under dust storm conditions. Time-dependent surface concentration and breathing zone exposure isopleths, evolving from a small contaminated area, show the potential hazard from wind eroding toxic materials. (auth)
Date: August 1, 1975
Creator: Travis, J.R.
Partner: UNT Libraries Government Documents Department

Monte Carlo simulation of the turbulent transport of airborne contaminants

Description: A generalized, three-dimensional Monte Carlo model and computer code (SPOOR) are described for simulating atmospheric transport and dispersal of small pollutant clouds. A cloud is represented by a large number of particles that we track by statistically sampling simulated wind and turbulence fields. These fields are based on generalized wind data for large-scale flow and turbulent energy spectra for the micro- and mesoscales. The large-scale field can be input from a climatological data base, or by means of real-time analyses, or from a separate, subjectively defined data base. We introduce the micro- and mesoscale wind fluctuations through a power spectral density, to include effects from a broad spectrum of turbulent-energy scales. The role of turbulence is simulated in both meander and dispersal. Complex flow fields and time-dependent diffusion rates are accounted for naturally, and shear effects are simulated automatically in the ensemble of particle trajectories. An important adjunct has been the development of computer-graphics displays. These include two- and three- dimensional (perspective) snapshots and color motion pictures of particle ensembles, plus running displays of differential and integral cloud characteristics. The model's versatility makes it a valuable atmospheric research tool that we can adapt easily into broader, multicomponent systems- analysis codes. Removal, transformation, dry or wet deposition, and resuspension of contaminant particles can be readily included. (auth)
Date: September 1, 1975
Creator: Watson, C.W. & Barr, S.
Partner: UNT Libraries Government Documents Department

Development program for the high-temperature nuclear process heat system

Description: A comprehensive development program plan for a high-temperature nuclear process heat system with a very high temperature gas-cooled reactor heat source is presented. The system would provide an interim substitute for fossil-fired sources and ultimately the vehicle for the production of substitute and synthetic fuels to replace petroleum and natural gas. The dwindling domestic reserves of petroleum and natural gas dictate major increases in the utilization of coal and nuclear sources to meet the national energy demand. The nuclear process heat system has significant potential in a unique combination of the two sources that is environmentally and economically attractive and technically sound: the production of synthetic fuels from coal. In the longer term, it could be the key component in hydrogen production from water processes that offer a substitute fuel and chemical feedstock free of dependence on fossil-fuel reserves. The proposed development program is threefold: a process studies program, a demonstration plant program, and a supportive research and development program. Optional development scenarios are presented and evaluated, and a selection is proposed and qualified. The interdependence of the three major program elements is examined, but particular emphasis is placed on the supportive research and development activities. A detailed description of proposed activities in the supportive research and development program with tentative costs and schedules is presented as an appendix with an assessment of current status and planning. (auth)
Date: September 1, 1975
Creator: Jiacoletti, R.J.
Partner: UNT Libraries Government Documents Department

Accurate determination of impurity concentrations in plutonium metals by statistical evaluation of analytical data

Description: Analytical data from a plutonium-metal exchange program conducted by six ERDA laboratories are statistically evaluated. The objective is an accurate determination of five metal impurities (aluminum, chromium, iron, nickel, silicon) in each of three plutonium metals by using data from four analytical methods. The statistical evaluation yields the weighted mean and its standard deviation for each method, plutonium metal, and impurity, using a procedure that minimizes the effect of outliers by assigning zero weights to the most extreme values and variable weights to the remaining data. Where possible, weighted means from the various analytical methods are pooled. (auth)
Date: July 1975
Creator: Martell, C. J.; Tietjen, G. L. & Horita, M. M.
Partner: UNT Libraries Government Documents Department