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PATHFINDER ATOMIC POWER PLANT STEAM SEPARATOR DEVELOPMENT
Development of a steam separator the Pathfinder Reactor is reported. A full-scale separator model was developed through the combination of scale-model testing and the application of principles associated with the existing theory of centrifugal separation. This model was put through full-scale air-water tests which led to modifications and a final design which meets Pathfinder requirements. Design data are included for the reactor and the steam separator. (J.R.D.)
Date:
June 15, 1962
Creator:
Kutsch, G. C.; Swanson, D. H. & Yant, H. W.
System:
The UNT Digital Library
THE DEVELOPMENT OF A SUMP-TYPE SOLIDIFIED-METAL SEAL
A solidified-metal seal for possible use in moIten-saIt systems was fabricated and tested on a laboratory scale/su The seal consisted of an 80 Au-20 Cu (wt%) sealant alloy in contact with IN0R-8 base metal/su Eleven successive helium-leak-tlght sealings were effected before termination of the test due to a leak in one of the mating parts/su With better control over the heating cycle and slight modifications in seal design, it is expected that the useful life of a seal of this type could be extended even further. (auth)
Date:
March 15, 1962
Creator:
Donnelly, R.G.
System:
The UNT Digital Library
Snap Shield Test Experiment Reactor Physics Tests
The initial physics tests on the Shield Test Experiment reactor and the precriticality rod-drop test data are presented. (auth)
Date:
July 15, 1962
Creator:
Tomlinson, R. L.; Johnson, R. P. & Wogulis, S. G.
System:
The UNT Digital Library
Emission Characteristics of Tantalum, Tungsten, Rhenium, and Iridium in Plasma Diodes
Experimental determinations of the ionic and electronic emission characteristics of Ta, W, Re, and lr cathodes in vapor thermionic converters are compared. It is shown that Ta provides superior thermal ionization qualities at high pressure compared with W, Re, and Lr. High electronic current densities may be obtained from Cs on Re and Cs on Ir at much lower Cs vapor pressures than from Ta or W. An over-all efficiency of 19% was achieved with a Re cathode at 2440 deg K. (auth)
Date:
March 15, 1962
Creator:
Gust, W. H.
System:
The UNT Digital Library
Reactor Development Program Progress Report, August 1961
Progress is reviewed on the following reactors: EBWR; Borax-V; ZPR-III- ZPR-VI; ZPR-IX; EBR-I; and EBR-II. An outline of fast and slow reactor safety studies in TREAT is presented. Progress is also reported in applied nuclear and reactor physics; development of reactor fuels, materials, and components; heat engineering technology; separation processes; and advanced reactor concepts. (T.F.H.)
Date:
September 15, 1961
Creator:
unknown
System:
The UNT Digital Library
Sodium Fluozirconate Precipitation Process for Zirconium Fuels. Part 1. Laboratory Development
Precipitation, evaporation, and extraction feed preparation conditions are established for the removal of zirconium and fluoride from fuel dissolver product solutions by the addition of sodium formate. A sparingly soluble complex fluozirconate is formed. Ninety-five to 99% of the zirconium and fluoride is separated from the uranium losses of 0.1% or less. Chemical material balances, based on experimental data, were developed for two flowsheets. In one flowsheet, sufficient nitric acid is added to the combined wash solution and filtrate produced during the precipitation step to destroy the formate ion (which inhibits uranium extraction) and to prevent post-precipitation during the evaporation of these solutions. The other flowsheet calls for addition of sufficient nitric acid to destroy the formate ion, but not enough to prevent post- precipitation during the concentration step. Post-precipitation removes additional zirconium and fluoride, but necessitates an additional solids- separation step. (auth)
Date:
May 15, 1962
Creator:
Newby, B. J.
System:
The UNT Digital Library
CONCEPTUAL DESIGN AND ECONOMIC EVALUATION OF A STEAM-COOLED FAST BREEDER REACTOR
A conceptual design and economic evaluation of 300 and 40 MW/.sub e/ steam-cooled fast breeder reactor power plants were performed. A reactor core composed of U-Pu oxide rod-type fuel elements clad with Inconel-X and surrounded by a blanket of depleted UO/sub 2/ fuel was studied in some detail. Reactor breeding ratios of from 1.27 to 1.5 and overall system doubling times of from 20 to 30 years are achievable. For the near term (1967) 300 MW/sub e/ plant, an energy cost of 7.6 mills/kwh is estimated, based on AEC ground rules for privately financed plants and utilities. This cost may go down to 5.7 mills/kwh by 1975. For the 40 MW/sub e/ plant corresponding energy costs are 19.5 and 13.7 mills/kwh, r -spectively. The R&D program required for this reactor concept is estimated at million with an additional million for improvements leading to the 1975 reactor. Investigation of the operational and safety aspects of the reactor indicated that satisfactory procedures can be used for startup, shutdown, and emergency cooling of the reactor. An increase in reactivity upon flooding can be prevented by incorprating small amounts of high resonance absorption material in the core. Preliminary calculations indicate a substantial increase in …
Date:
November 15, 1961
Creator:
Sofer, G.; Hankel, R.; Goldstein, L. & Birman, G.
System:
The UNT Digital Library
Reactor Development Program Progress Report, March 1962
ABS>Developmental work is reported on the EBWR and Borax-V other general development work is reported in the area of liquid metal cooled reactors and particularly on the EBR-I and H, and the fast reactor test facility. General reactor technological development is described on applied reactor physics, reactor fuels development, reactor materials development, heat engineering and fluid flow, chemical separations, advanced systems, and nuclear safety. (J.R.D.)
Date:
April 15, 1962
Creator:
unknown
System:
The UNT Digital Library
Reactor Development Program Progress Report (for) July 1961
A summary is presented of activities in reactor and general engineering research programs. Discussions are included for developments in EBWR, BORAX-V, ZPR-III. ZPR-VI, ZPR-IX, EBR-I, and EBR-II. Reactor safety studies were performed for fast and thermal reactors. Nuclear technology developments are discussed for applied nuclear and reactor physics, reactor fuels and materials development, heat engineering studies, separations processes, and advanced reactor concepts. (B.O.G.)
Date:
August 15, 1961
Creator:
unknown
System:
The UNT Digital Library
Exploration of the Planets: A History, A Prospect, and Some Issues
This report discusses the history and future of planetary exploration programs by the U.S. and the Soviet Union. Topics covered are motivations for planetary exploration, unmanned missions sent to Mars and Venus by the U.S. and Soviet Union, plans for manned missions to planets such as Mars, technical needs and costs for planetary space missions, and long-term plans and ideas.
Date:
December 15, 1969
Creator:
Devoe, Barbara M. & Sheldon, Charles S., II
System:
The UNT Digital Library
SALT PHASE CHLORINATION OF REACTOR FUELS. IV. NIOBIUM BEHAVIOR IN THE LEAD CHLORIDE AND CHLORINE-LEAD CHLORIDE SYSTEMS
Investigation of the behavior of Nb in PbCl/sub 2/ showed that the Nb dissolution rate is quite low. Since stirring increases the rate appreciably, it is likely that the initial reaction is diffusion controlled. A subsequent reaction is linear in rate with an activation energy of 23.4 kcal per mole. Incomplete volatilization of dissolved Nb from PbCl/sub 2/ is due to the presence of lower oxidation states of Nb. The addition of Cl to PbCl/sub 2/ increases the rate of dissolution of Nb. The rate is high enough at reasonable temperatures to be practical for fuel dissolution, e.g. 12 mg min/sup -1/ cm/sup -2/ at 550 deg C and 120 mg min/sup -1/ Cl. Volatilization of NbCl/sub 5/ is 99.9% complete at 550 deg C. (auth)
Date:
March 15, 1962
Creator:
Teague, J. L.; Hahn, H. T. & Vander Wall, E. M.
System:
The UNT Digital Library
GUIDE TO NUCLEAR POWER COST EVALUATION. VOLUME 4. FUEL CYCLE COSTS
Information on fuel cycle cost is presented. Topics covered include: nuclear fuel, fuel management, fuel cost, fissionable material cost, use charge, conversion and fabrication costs, processing cost, and shipping cost. (M.C.G.)
Date:
March 15, 1962
Creator:
unknown
System:
The UNT Digital Library
AN ANALYSIS TO DETERMINE THE PERCENTAGE OF HELIUM BYPASSING THE CORE DUE TO THE REFLECTOR SEALING SYSTEM DURING NORMAL OPERATION OF THE HTGR
The percentage of helium which will bypass the core if the reflector system shown is used is predicted. It is estimated that nominally about 0.1 to 0.2% of the total flow will bypass the core, which is not considered excessive. The most difficult parameter to determine was Z, the gap between the sealing surfaces. The method used to predict Z is presented. The effect of bowing due to a temperature gradient across the seals is discussed. (auth)
Date:
November 15, 1961
Creator:
Nimtz, F.B.
System:
The UNT Digital Library
VOID COEFFICIENT OF REACTIVITY ASSOCIATED WITH THE ISLAND REGION OF THE HFIR
Changes in neutron multiplication caused by voids in the island of the HFlR were calculated and measured experimentally. The results indicated that with only water initially in the island the maximum change in neutron multiplication ( DELTA k/sub max) associated with island voids is 0.032 with a corresponding void fraction of 70%. With a simulated 300 g Pu target in the island DELTA k/sub max/ was 0.0l6, and the corresponding void fraction was 42%. In view of these large changes in neutron multiplication, calculations were made to determine what additional materials could be used in the island to reduce DELTA k/sub max/ and what the associated decrease in peak thermal flux wouId be. The results indicated that of the materials considered the use of beryllium in the water island resulted in the smallest decrease in flux for a specified DELTA k/sub max/. To reduce DELTA k/sub max/ to 0.01 required 26% by volume of beryllium in the island; the corresponding reduction in thermal flux, as compared to an all-water island, was about 10%. In order to reduce DELTA k/sub max/ to 0.0l with a 300 g Pu target in the island, the aIuminum-to-water ratio of the target had to be …
Date:
November 15, 1961
Creator:
Cheverton, R.D.
System:
The UNT Digital Library
STEAM-COOLED POWER REACTOR EVALUATION, STEAM-COOLED FAST BREEDER REACTOR
Conceptual design and economic studies of a steamcooled fast breeder reactor that can also be used as a source of power are presented. Two reactor plant sizes were considered: a 300-Mw(e) central power station plant and a 40 Mw(e) plant. It was concluded that attractive economics and good breeding characteristics breeding ratios from 1.27 to 1.42) can be achieved in steam- cooled PuO/sub 2/UO/sub 2/ fueled fast reactors. Low capital costs can be obtained by a compact reactor core and the absence of large heat exchangers and complicated process systems. Reactor design data are discussed. Analysis showed that these reactors can be prevented from going prompt critical, when fully flooded, by incorporating a tolerable amount of high resonance absorption materials such as hafnium or indium. An increase in reactivity on loss of coolant was indicated by preliminary calculations. (M.C.G.)
Date:
April 15, 1961
Creator:
Sofer, G.; Hankel, R.; Goldstein, L. & Birman, G.
System:
The UNT Digital Library
SNAP II POWER CONVERSION SYSTEM TOPICAL REPORT NO. 11, ORBITAL FORCE FIELD BOILING AND CONDENSING EXPERIMENT
The characteristics of Rankine space power plants in the zero gravity aspect of the environment of space were lnvestigated. The expected effects of Rankine space power plants are described. Discussions of experimental techniques for studying these phenomena show that this information can be obtained rapidly and economically. Recommendations for a program to supplement SNAP II and slmllar Ranklne space power development efforts in this vital area are made, and consist of: the development and testing of a small system that adequately simulates a complete Ranklne system, first in zero grayity and finally, in the complete orbltal environment; followed by, the development and similar testing of a complete Rankine system using SNAP ll hardware. (auth)
Date:
January 15, 1960
Creator:
Grevstad, P.E.
System:
The UNT Digital Library
SNAP-23A Program, Phase I. Thermoelectric Converter Development Program. Monthly Report, February 1969.
None
Date:
March 15, 1969
Creator:
unknown
System:
The UNT Digital Library
FABRICATION OF CERAMIC INTERNAL REFLECTOR FOR THE SNAP 8 EXPERIMENTAL REACTOR
Fabrication of internal reflector pieces for the SNAP-8 core is described. These reflectors were made of BeO, with and without the addition of Sm/sub 2/O/sub 3/ as a nuclear poison. Because of the high density and dimensional tolerance requirements, the complexity of shapes, and the comparatively modest number of parts to be produced (approximately 1000), the blanks were hot pressed and subsequently machined with diamond wheels and cores. In almost all cases, the specified density of 98% of theoretical was achieved. The low eutectic temperature in the BeO--Sm/sub 2/O/sub 3/ system of about 1420 deg C necessitated special pressing parameters, which were arrived at by experimentation. A rather coarse grit size (80) was used for the diamond wheels, which resulted in very little wear to the wheels, and an extremely low dimensional rejection rate on the blanks. Because of the toxicity of BeO, all equipment was enclosed, and was held under negative pressure. (auth)
Date:
July 15, 1962
Creator:
Langrod, K.
System:
The UNT Digital Library
Reactor Development Program Progress Report, September 1961
BS>Data from examination of blade-type control rods which were used in BORAX are discussed. Operation and maintenance of EBWR is outlined. In work on Borax V, modifications for easier installation of reactor and components is outlined followed by discussion of superheat fuel element development, and fabrication of various reactor components. Borax reactor design is also reported along with information on development and testing. In research on sodiumcooled reactors, activities are summarized in the LPR III and LPR IV programs along with developmental work on EBR I and II. Studies on reactor safety are reported and activities in a program of nuclear technology and general support are outlined. (J.R.D.)
Date:
October 15, 1961
Creator:
unknown
System:
The UNT Digital Library
DEVELOPMENT OF ULTRASONIC TECHNIQUES FOR THE REMOTE MEASUREMENT OF THE HRT CORE VESSEL WALL THICKNESS
Design and development of a remote ultrasonic inspection technique for use in measuring wall thicknesses in the HRT core vessel are described. (J.R.D.)
Date:
March 15, 1962
Creator:
McClung, R.W. & Cook, K.V.
System:
The UNT Digital Library
Spert I Destructive Test Program Safety Analysis Report
The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor period range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation …
Date:
June 15, 1962
Creator:
Spano, A. H. & Miller, R. W.
System:
The UNT Digital Library
DETERMINATIONS OF THE KINETICS AND MECHANISMS OF DEBORONIZATION AT 1135 C
The mechanisms and kinetics of the loss of boron during heating at 1135 deg C under various dynamic environments were determined from powder compacts of 5 wt% elemental boron dispersed in matrices of Fe, Cr, Ni, Si, Fe/sub 2/O/sub 3/, Cr/sub 2/O/sub 3/, NiO, and SiO/sub 2/, compacts of austenitic stainless steel alloy powder containing 0.25 wt% boron, and wrought specimens of 0.13 wt% boron-- stainless steel alloy. The compacts containing 5 wt% boron were heat treated in vacuum, highpurity argon, wet helium, and hydrogen. With the exception of those heat treated in hydrogen, significant boron losses occurred only when a supply of oxygen, either from the sample itself or as a deliberate addition to the heat- treating environment, was available. Correspondingly, the loss mechanism is postulated to be the oxidation of boron to boron sesquioxide and its volatilization from the sample. The loss rate is controlled by the volatilization rate of the oxide which is directly influenced by structure of the compact and sintering environment. Independent of the chemical nature of the matrix, boron losses were incurred during heat treatment in hydrogen. Variations of the water content of the hydrogen from 7 to 460 ppm did not significantly influence …
Date:
September 15, 1961
Creator:
Cherubini, J.H.
System:
The UNT Digital Library
Safety Calculations for MsSRE
A number of conceivable reactivity accidents were analyzed, using conservatively pessimistic assumptions and approximations, to permit evaluation of reactor safety. Most of the calculations, which are described in detail, were performed by a digital kinetics program, MURGATROYD. Some analog analyses were also made. None of the accidents which were analyzed lead to catastrophic failure of the reactor, which is the primary consideration. Some internal damage to the reactor from undesirably high temperatures could result from extreme cold- slug accidents, premature criticality during filling, or uncontrolled rod withdrawal. Each of these accidents could happen only by compounded failure of protective devices, and in each case there exists means of effective corrective action independent of the primary protection, so that damage is unlikeIy. The calculated response to arbitrary ramp and step additions of reactivity show that damaging pressures could occur only if the addition is the equivalent of a step of about 1% delta k/k or greater. (auth)
Date:
May 15, 1962
Creator:
Haubenreich, P.N. & Engel, J.R.
System:
The UNT Digital Library
Reactor Development Program Progress Report, September 1962
Progress is reported on development of liquid metal and water cooled reactors. Details are included concerning the EBWR, Borax V, ZPR-III, ZPR-VI, AFSR, EBR-I, EBR-II, and FARET. Developments in general reactor technology are reported in applied reactor physics, reactor fuel and components development, heat engineering, and chemical separations. Advanced systems research and development was devoted to the Argonne Advanced Research Reactor and a conduction- cooled reactor to substitute for isotope heat sources. Safety studies are reported on thermal and fast reactors. (J.R.D.)
Date:
October 15, 1962
Creator:
unknown
System:
The UNT Digital Library