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0-2 kv Flash Tube Supplies
The power supplies designed and constructed to power high-intensity flash tubes are described. Three supplies, capable of charging 100 mfd to 2 kv with a repetltion rate of not less than 0.8 sec, are operated remotely from a control panel containing 3 powerstats. The power supplies are full wave center trapped rectifiers employing silicon rectifiers. (M.C.G.)
T = 0 AND T = 1 PAIRING IN LIGHT NUCLEI.
No Description Available.
1.5% boron, stainless steel
No Description Available.
A 1.5 Mwe Thermionic Reactor Space Power System
No Description Available.
E-1 common analog model
No Description Available.
04 nuclear safety: pressure piping crack monitoring detection of metal overstress by acoustic emission. Progress report, July-September 1966
The three main areas of effort have been: (1) definition of the general acoustic response pattern related to the gross aspects of forming and extending a crack in various materials, (2) development of a monitor system prototype concept exclusive of transducers and (3) development of a suitable, high temperature transducer. Tests using double cantilever beam (DCB) specimens of various materials to establish conditions of crack formation and growth have indicated that material ductility is a major controlling factor in the acoustic response pattern. It appears to effect both acoustic emission intensity and the point in the crack formation-growth sequence at which the main emission occurs. A concept has been developed for the prototype of a full scale monitor system. Hardware development is being limited to the analyzer portion of the system at this time because it is the part most significant to demonstrating feasibility of the intended application. Signal level and signal rate are both being investigated as possible parameters for evaluating acoustic emission data. Of the various transducers for potential high temperature application, the capacitive or electrostatic transducer now looks most promising. A significant improvement in sensitivity has been achieved and a trial model used during recent tests produced generally satisfactory data. The sequence of effort on the program is being adjusted somewhat from that previously outlined. Some of the more detailed investigative phases will receive only moderate attention, temporarily, in favor of first demonstrating the basic feasibility of detecting acoustic emission and making a meaningful analysis under postulated service conditions.
5 and 5 hot cell configuration for E-MAD facility. Phase II
No Description Available.
A 5 mechanical test of core support block. Preliminary report
No Description Available.
6 kv CAPACITOR CHARGING SUPPLY
The power supplies designed and constructed to power high intensity flash tubes used in bubble chamber experiments are briefly described and are accompanied by a schematic diagram of the layout. (D.C.W.)
9-Zoom : A One-Dimensional, Multigroup, Neutron Diffusion Theory Reactor Code for the IBM 709
The following document describes the usage of the LRL 9-ZOOM code, a neutron diffusion theory reactor code for the IBM 709. The code has been modified to solve configuration of a series of stacked cylindrical disks, designating a new geometry case.
10-30 Bev/c ELASTIC SCATTERING OF $pi$$sup +-$+ p, p + p, /anti p/ + p AND K$sup +-$ + p OVER THE /t/ RANGE 0.0005 TO 1 (Bev/c)$sup 2$
No Description Available.
A-11 seven cluster model: FFL-10 solid core tests
No Description Available.
A-11 seven cluster model: Phase V, flow induced vibration tests
No Description Available.
12-INCH SODIUM FLOW CONTROLLER. Technical Manual 20357
A manual is presented for the Sodium Flow Controller used in controlling flow to regulate heat transfer in a liquid metal nuclear power plant. A description of the controller, general installation and operational pointers, installation instructions, instructions for dismantling of the Sodium Flow Controller, instructions for assembly of Sodium Flow Controller, list of special tools and fixtures, and repair parts list are given. (M.C.G.)
12" SODIUM FLOW CONTROLLER. PERMANENT MAGNET COUPLING. MAGNETIC CALCULATION MANUFACTURE AND TEST RESULTS
In order to retain the hermetic feature of the sodium flow controller, magnetic flux linkage of a permanent magnet coupling is used to transmit the torque produced by the operator through the pressure wall. Magnetic calculations, manufacture, and testing of this magnet coupling are described. (M.C.G.)
13-Watt Curium Fueled Thermoelectric Generator for a Six-Month Space Mission. Final Report
No Description Available.
13-Watt Curium-Fueled Thermoelectric Generator for Hard Lunar Impact Mission. Final Report-Subtask 5.8
Results of a conceptual design study for a curium powered thermoelectric generator of minimum size and weight which is capable of sustaining hard impact is presented. The generator produces a minimum of 13 watts of d-c power at 3 volts, and weighs 6.2 pounds excluding shielding. (J.R.D.)
14-Inch Swing Check Valve Test
The check valve for the Hallam Power Reactor uses a knife-edge bearing for the flapper in place of the usual journal-type bearing. Mechanical cycling in sodium at 600 deg F was used to check operation of this bearing. A total of 309 mechanical cycles was completed with no apparent malfunctioning of the valve. Measured leskage rates were 0.46 gpm at 0.93 psig, 0.73 gpm at 3.4 psig. and 0.32 gpm at 5.9 psig. (M.C.G.)
20-ton and 1/2-ton High Explosive Cratering Experiments in Basalt Rock: Final Report, August 1962
From preface: Project Buckboard, a Plowshare sponsored effort, involved the detonation of three 40,000-pound and ten 1000-pound high-explosive charges in basalt rock.
20-ton HE Cratering Experiments in Desert Alluvium: Final Report, May 1962
From abstract and summary: Project Stagecoach consisted of the detonation of three 40,000-pound charges. Blocks of cast TNT were stacked to resemble a sphere and, the whole center-detonated.
A 23-Group Neutron Thermalization Cross Section Library
A set of 23-group neutron cross sections for use in the calculation of neutron thermalization and thermal neutron spectral effects in SNAP reactors is compiled. The sources and methods used to obtain the cross sections are described. (auth)
The 25-Inch Liquid Hydrogen Bubble Chamber
No Description Available.
30 Megawatt Heat Exchanger and Steam Generator for Sodium Cooled Reactor System
Final design for the 30 megawatt intermediate heat exchanger and steam generator.
30 Megawatt Heat Exchanger and Steam Generator for Sodium Cooled Reactor System: Volume 2, Chemical and Stress Analysis
Chemical engineering analysis and stress analysis for design of the 3 megawatt intermediate heat exchanger and steam generator for service with a liquid sodium heat transfer fluid.
30 Megawatt Heat Exchanger and Steam Generator for Sodium Cooled Reactor System: Volume 4, Operation and Maintenance Procedures
Operation and maintenance procedures for 30 megawatt heat exchanger and steam generator for sodium cooled reactor system.
40-Mw(e) Prototype High-Temperature Gas-Cooled Reactor Postconstruction Research and Development Program. Quarterly Progress Report for the Period Ending April 30, 1965
No Description Available.
40-Mw(e) Prototype High-Temperature Gas-Cooled Reactor Postconstruction Research and Development Program. Summary Progress Report for the Period Ending, October 1964
No Description Available.
40-MW(e) Prototype High-Temperature Gas-Cooled Reactor Research and Development Program. Quarterly Progress Report for the Period Ending June 30, 1962
Research and development progress specifically directed toward the construction of a 40-Mw(e) prototype power plant employing a high-temperature, gas-cooled, graphitemoderated reactor known as the HTGR is reported. Irradiation of element III-B in the in-pile loop continued satisfactorily. The element has generated a total of l36.3 Mw-hr of fission heat. The gross activity in the purge stream increased slightly to about 350 mu C/cm/sup 3/. By taking larger gas samples than were previously taken, a value of 0.02 VC/cm/sup 3/ was obtained for the gross activity of the primary loop. Element III-A, which was removed from the loop after generating 133 Mw-hr of fission heat, was disassembled and examined. No fuel-compact damage of any type was visible. Determination of the distribution of fission products in the element is under way, Fissionproduct- release data for in-pile-loop element III-A were calculated. During the 133 Mw- hr of operation, the release fraction increased by approximately one order of magnitude. Also calculated were the xenon and krypton release data for the first 100 Mw-hr of III-B operation. The release rate for the longer-lived isotopes increased bv about a factor of 10 and that of the shorter-lived isotopes by about a factor of 100. A test was run in which the in-pileloop purge flow, was stopped. The primariy-loop activity level rose sharply during the first hour, increased at a slower rate for the next 11 hr, and then appeared to level off. When purge flow was resumed, the gross activity in the primary loop was cleaned up with a half life of about 2.2 hr. An attempt was made to identify Cs/sup 137/ and Ba/ sup 140/ plateout in portions of the in-pile loop. A very small amount of cesium (less than a monolayer) was found, but no barium could be detected. The validity of two ...
40-Mw(E) Prototype High-Temperature Gas-Cooled Reactor. Research and Development Program. Quarterly Progress Report for the Period Ending March 31, 1963
No Description Available.
40-MW(E) PROTOTYPE HIGH-TEMPERATURE GAS-COOLED REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Summary Report for the Period January 1, 1959-December 31, 1959 and Quarterly Progress Report for the Period October 1, 1959-December 31, 1959
The HTGR prototype plant (Peach Bottom Power Reactor) is being designed to produce steam at l450 psi and 1000 deg F and to have a net capacity of 40 Mw(e). The fuel temperatures and gas pressures will be approximately the same as those required for larger plants. The reactor data and operating conditions for the graphite-clad core are given. The reactor and primary coolant systems are described. The prospects for development of the graphite-clad fuel element in time for use in the first loading of the reactor were improved by important advances in methods of fabrication and testing of both fuel compacts and graphite sleeves. The hot-pressing process for making fuel compacts was used successfully to make full-size compacts with a uniform distribution of ThC/sub 2/- UC/sub 2/ particles. Three irradiation capsules were fabricated and inserted in a test reactor to determine fuel compact and sleeve performance under HTGR conditions of irradiation and temperature. Two of these ran satisfactorily for the scheduled time of operation. A scope design study of the in-pile loop that will be used to evaluate the full-diameter graphite-clad element was completed. Experiments to determine the extent of fuel migration within the element were undertaken. Preliminary results indicated that the central fuel-element temperatures must not exceed 2300-C for routine operation. An important start was made in developing an understanding of how to treat the neutron thermalization process in high-temperature graphite reactors. Analytical techniques for calculating the thermal neutron spectra in poisoned graphite media were developed and programmed for the IBM 704 computer. The experimental technique of measuring neutron spectra by using a pulsed linear electron accelerator was demonstrated by measurements made with boron-loaded graphite. A mockup of a small portion of the reactor core was constructed and operated to determine the local heat-transfer coefficients and pressure drop ...
40-MW(E) PROTOTYPE HIGHTEMPERATURE GAS-COOLED REACTOR. POST CONSTRUCTION RESEARCH AND DEVELOPMENT PROGRAM. Quarterly Progress Report for Period Ending January 31, 1965
No Description Available.
40-tube overbore facility location, C Reactor
Possible locations of the projected 40-tube overbore facility at the C Reactor are discussed from the standpoint of obtaining conversion ratio data applicable to a full-reactor overbore program.
75,000 KILOWATTS OF ELECTRICITY BY NUCLEAR FISSION AT THE HALLAM NUCLEAR POWER FACILITY
For presentation at ASCE Convention in Reno, Nevada on Thursday, June 23, 1860. A description of the Hallam Nuclear Power Facslity is presented. The history of the project, program participants, site description, component development program, reaetor building, reactor structure, reactor core, sodium systems, instrumentation and control, fuel and component handling, auxsilary sustems, special design features, and advantages of sodium graphite reactor systems are discussed. (M.C.G.)
A 90$sup 0$ $sup 3$He NEUTRON SPECTROMETER.
No Description Available.
100-C water plant
System curves for each portion of the C Area Water Plant were obtained from referenced work and are presented in figures. Field test data, corroborating the calculated curves, are presented as singular points on the same graphs. Present maxima capacity of the C Area Filter Plant was 121,000 gpm with 118,000 gpm available for use as primary reactor coolant. Modifications to the filter effluent piping would increase this available flow to about 180,000 gpm. Of the 118,000 gpm available for C Reactor use, 10,000 to 12,000 gpm was demanded by B Area through the 183 BC intertie. The maximum flow that the intertie line could handle, without reducing the filter capacity of the C Area filters, is about 21,000 gpm.
100-K Area Downcomer Test Data Project CGI-883
105-KE downcomer pressure data are tabulated.
100-K Area electrical power system load and voltage study for project CG-775. Revision
The proposed increased water capacity for 100-K plants will increase the electrical load to be supplied. The load study showed that the capacity of the existing 13.8 kV system is adequate to carry the increased loads proposed for Project CG-775, while for the 5 kV system, an expanded power system is proposed. Likewise, the voltage regulation on the kV system bus will be excessive, and voltage regulators should be added.
100-KEW coolant backup adequacy
Short communication.
100-N technical manual. Volume 2A: Systems descriptions
This report contains engineering drawings for the control room, reactor monitoring systems, and reactor control systems for the N reactor. Each console in the control room is detailed. Other systems discussed are: stack air monitoring system, charging machine control systems, and heating and ventilation control systems. A N reactor plant glossary is included.
100-Watt Curium-242 Fueled Thermoelectric Generator--Conceptual Design. SNAP Subtask 5.7 Final Report
A thermoelectric generator which produces 100 watts of electrical power continuously over a six-month operational life in a space environment was designed. It employs the heat produced by the decay of Cm/sup 24/ as the source of power. Uniform output over the operational life of the generator is accomplished by means of a thermally actuated shutter which maintains the hot junction temperature of the thermoelectric conventer at a constunt figure by varying the amount of surplus heat which is radiated directly to space from the heat source. The isotopic heat source is designed to safely contain the Cm/sup 242/ under conditions of launch pad abont and rocket failure, but to burn up upon re-entry to the earth's atmosphere from orbital velocity. (W.L.H.)
105-C overbore 40 tube test process tube assembly flow and pressure drop calibration test
The object of this test is to determine the hydraulic characteristics of the proposed overbore process tube assembly designs which are to be installed on 105-C reactor for the 40 tube overbore fuel element test.
105-C overboring thirteen tube outage, March 6, 1961--March 10, 1961
C Reactor was shut down on a scheduled basis at 8:30 a.m. March 6, 1961 for the purpose of overboring 17 process channels. this report will cover that outage and discuss problems encountered in completing the tasks involved in overboring.
105-F intermediate range test hole parameters
This report discusses 105 intermediate range test hole which was bored in 105-F to demonstrate reactor hole boring techniques. This test hole was then utilized to determine the performance of neutron sensitive gamma compensated ion chambers (CIC), the effectiveness of epoxy reactor shield plug designs, the compatability of the prototype.components intermediate range instrumentation, and demonstrate the performance of an intermediate range monitor system on a typical Hanford reactor application. In the process of determining the above information physical, nuclear, gamma, and temperature data was recorded as related to the 105-F intermediate range monitor test hole and is provided in this report.
I-131 in milk from Cabriolet fallout
No abstract available.
190-DR steam turbine backup adequacy report
Because of the questionable performance of the 190-DR steam backup system during the power outage of April 6, 1962, it is felt that a general review of the DR secondary water system capabilities and the standby status of the pumping units is warranted. This report shall briefly describe the performance of the steam backup system during the April 6 power outage and the subsequent power outage of July 10, 1962. Since the former outage, tests have been conducted on the steam pump units to determine their capabilities; the test results are presented in this report. A statement of the generally accepted criteria for secondary coolant system adequacy is included and recommendations for meeting the criteria at DR reactor are presented.
190-H drawdown test
A discrepancy of about 1000 gpm has existed between the full-flow recorded 190, 105 and ROL flows. While past operating practices have not used the 190 or ROL flow rates for official purposes, the disquieting, though not theoretically unexplicable, differences require some quantitative resolution. On November 24, 1962, a drawdown test of the 190-H storage tanks was performed to establish the accuracy of the various flowmeters. The drawdown test of the 190 storage tanks was run at the beginning of a scheduled reactor shutdown. With the full reactor flow supplied by the electric process pumps feeding from the storage tanks, the 183-H supply to the storage tanks was valved off. Additionally, non-process water usually taken from the storage tanks was valved off. The storage tank water levels were taken, then recorded as a function of time.
200 Area monthly report, August 1966
This report details activities of the 200 Area for the month of August 1966.
200 Area monthly report, July 1966
This report details 200 Area activities for the month of July 1966.
200 Area monthly report, June 1966
No Description Available.
200 Area monthly report, May 1966. Report No. 5
This report details 200 Area activities for the month of May 1966.