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NACA Research Memorandums
159
Atomic Energy Commission Reports
94
Hanford Works Reports
52
HW (Series)
40
Personal Papers (MS 80-0002)
32
North American Aviation Reports
24
AEC research and development report
20
BMI (Series)
19
Battelle Memorial Institute Reports
19
Hanford Laboratories Operation Monthly Activities Report
17
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Analysis of Radiation From Hnpf Cold Traps and Primary Sodium Pumps During Removal and Shipping
The expected maximum contamination of the HNPF cold traps and primary sodium pumps was determined along with the maximum dose rates from these components during removal and shipping. Suitable shielding for casks to be used in the removal operation and for shipping these components away from the reactor site is specified. Access to an unshielded cold trap is limited by high dose rates, i.e., 100 mr/hr at 120 ft, after 180 days decay time. A handling cask providing a radial shield of 3 in. of lead will provide adequate personnel protection for the removal operation, if 180 days decay time is allowed before the trap is removed. An additional 2.4 in. of lead is required for offsite shipment of the cask. This additional shielding can be added after the trap is removed from the reactor building. Dose rates from the cold trap after the shield plug is removed from the access hole are shown. If direct line-ofsight exposure is avoided, dose rates to personnel will be below 100 mr/hr at any position, and below 10 mr/hr at distances greater than 20 ft from the access hole. Dose rates from the cask during its travel away from the hole, will be …
Date:
December 15, 1959
Creator:
Rhoades, W. A.
System:
The UNT Digital Library
The Early Antiproton Work [Nobel Lecture]
Early work on the antiproton, particularly that part which led to the first paper on the subject, is described. Conclusions that can be drawn purely from the existence of the antiproton are discussed. (W.D.M.)
Date:
December 15, 1959
Creator:
Chamberlain, O.
System:
The UNT Digital Library
Fission Gas Generation and Consequent High Temperature Irradiation Effects on Natural Uranium, Including Creep Properties : a Literature Search
Sixty-two references, some annotated, were taken from foreign literature to determine the status of work done in this area of metallurgy during 1958 and 1959. The references are grouped into two categories: references pertinent to the subject and general metallurgical references on metallic fuels.
Date:
December 15, 1959
Creator:
Scott, Myra
System:
The UNT Digital Library
Hanford Laboratories Operation Monthly Activities Report: November 1959
This is the monthly report for the Hanford Laboratories Operation, November 1959. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.
Date:
December 15, 1959
Creator:
Hanford Laboratories
System:
The UNT Digital Library
Hazards Analysis of the Organic Moderated Reactor Experiment
Introduction: The description of the Organic Moderated Reactor Experiment, (OMRE), its location, its safety system, and operative procedures have been previously detailed. The present report, although dealing with the subject of OMRE safety, has the more detailed intent of (1) determining the behavior of the OMRE under extremely unlikely sets of conditions; and (2) providing additional design information in the areas of reactivity coefficients, burnout heat flux, and reactor control.
Date:
December 15, 1959
Creator:
Williams, R. O., Jr.; Allen, W. O.; Ash, E. B.; Scott, W. W.; Shimazaki, T. T.; Sletten, H. L. et al.
System:
The UNT Digital Library
HNPF Cold Trap Evaluation
Two designs of sodium cold traps for the HNPF have been subjected to full scale tests, Performance features that were investigated include oxide removal efficiency, oxide capacity, pressure drop characteristics, economizer effectiveness, and temperature profiles, Results indicate that both designs should perform satisfactorily in the Hallam plant, (auth)
Date:
December 15, 1959
Creator:
Cygan, R.
System:
The UNT Digital Library
HRT Process Flowsheets--Revised Edition
Revised HRT flowsheets are presented. These revisions cover such items as relocation of freezer units on the lines, corrections to the numbering of lines, valves or instruments, and the addition of a few lines in the service areas. The waste and vent system flowsheet was redrawn as two sheets. (C.J.G.)
Date:
December 15, 1959
Creator:
Robertson, R. C. & Jones, J. E.
System:
The UNT Digital Library
IONIUM (THORIUM-230) FOR RADIOISOTOPE PREPARATION--STATUS REPORT
The general prospects of several radioisotopes are reviewed; the special properties of U/sup 232/ and Th/sup 228/ are poi nted out; and ionium (Th/sup 230/ ) and protactinium target materials are discussed from the sthndpoint of availability and chemical separations processes required for the preparation of U/ sup 232/ and Th/sup 228/. Outlines are given for potential schem es for the separation of U/sup 232/ and Th/sup 228/ from uranium milling pr ocess waste streams and from the irradiation products of Th/sup 230/--Th/sup 232/ mixtures. The high heat generating rates of these potent alpha emitters make them especially suitable for primary consideration as heat sources for small thermoelectric generators. The exceptionally high alpha activity suggests their use in special neutron sources as Ra-Be sources, and they may have sufficiently high neutron generating rates to be in contention with some of the smaller research reactors and experimental neutron producers. (B.O.G.)
Date:
December 15, 1959
Creator:
Coppinger, E.A. & Rohrmann, C.A.
System:
The UNT Digital Library
Ionium (Thorium-230) for Radioisotope Preparation (Status Report)
For many years effort has been directed toward the development of means for the practical utilization of the heat evolved from the radioactive decay of certain fission and irradiation products. Fission products, in view of their availability in the plentiful, high-level wastes resulting from the processing of irradiated, nuclear reactor fuels have been most intensively studied for applications where their heat of decay might be converted into useful energy such as electricity for the operations of certain devices in place of chemical batteries . In addition other materials having desired radioactive properties may be produced by neutron bombardment of readily available elements.
Date:
December 15, 1959
Creator:
Coppinger, E. A. & Rohrmann, C. A.
System:
The UNT Digital Library
IRRADIATION EFFECTS ON THE SURFACE REACTIONS OF METALS. Summary Report for October 1, 1958 to November 1, 1959
Weight increases during the oxidation of irradiated foils of pure copper were greater than for unirraaiated specimens. Enhanced reactivity appeared to be strongest in the thin-film region up to about 5 mu g/cm/sub 2/. Oxide film (Cu/ sub 2/O) thickness for both irradiated and unirradiated specimens was approximately 1200 A. Radiation did not affect the reduction of Cu/sub 2/O during the induction period (period in which the reduction proceeds very slowly or not at all). In later stages of the reduction process, a serious lack of reproducibility was observed. Radiation effects on films of Cu/sub 2/O formed by prior oxidation of the copper substrate decreased the kinetics of secondary oxidation. The secondary oxidation curve exhibited a large gap at the point of interrnption for irradiation. The development of an automatic recording microbalance of high sensitivity and a furnace for studies in reactor radiation fields is reported. Measurements were made of the electrode potentials of irradiated (5.5 x 10/sup 19/ neutrons cm/sup -2/) copper, aluminum, magnesium, and zirconium. Cell potentials were found to be dominated by the oxide films formed on the electrode surfaces. The results indicate that radiation does affect the local anode reaction potential. No significant difference between the …
Date:
December 15, 1959
Creator:
Carpenter, F. D. & White, J. L.
System:
The UNT Digital Library
Nuclear Powered Tanker Design and Economic Analysis - Direct Cycle Boiling Water Reactor
A design and economic analysis is presented for a 44,000 DWT tanker utilizing a 27,300 shp direct cycle boiling water reactor propulsion system. Included are design descriptions and drawings, estimates of operating and capitol costs for the reference ship, estimates of long range potential, descriptions of required research and development programs, a preliminary safeguards analysis, an analysis of ship's motion on boiling water reactor performance, a summary reactor physics analysis, and a discussion of direct cycle accessibility considerations. Also included are summary descriptions for various design alternatives which were considered.
Date:
December 15, 1959
Creator:
General Electric Company
System:
The UNT Digital Library
Power Flattening in Sodium Graphite Reactors by Spatial Variation of Moderator Properties
Abstract: In the present study, the variation of moderator composition was postulated to be effected by the inclusion of varying amounts of beryllium oxide in the graphite of an SGR.
Date:
December 15, 1959
Creator:
Connolly, T. J.
System:
The UNT Digital Library
Studies of Nuclear Debris in Precipitation : Second Quarterly Progress Report
A sequential method is described for the determination of strontium- 89, strontium-90, barium-140, and tungsten185 in rain water samples. A comparison is made of data on radioactivity in individual rain samples and in monthly pot samples. Data are included on the total strontiuro-90 fall-out collected in steep walled stainless steel pots during the period. Results are compared with cumulative data for 1958 end 1959. (C.H.)
Date:
December 15, 1959
Creator:
unknown
System:
The UNT Digital Library
Casting Development for Uranium-Molybdenum Alloy Shapes
The casting of shapes of uranium--molybdenum metal of varying sizes and thicknesses from a molten charge has been successfally accomplished with specificially designed graphite distributors and molds. Solid cylinders, hollow cylinders, and flat plate shapes were cast in gang molds. As many as 35 solid cylinders have been cast simultaneously. All castings had smooth surfaces, and solid shapes were cast to 0.006-in. tolerance on all dimensions except length. (auth)
Date:
November 15, 1959
Creator:
Binstock, M. H. & Stanley, J. A.
System:
The UNT Digital Library
FAST OXIDE BREEDER-REACTOR. PART I. PARAMETRIC STUDY OF 300(e) MW REACTOR CORE
Physics scoping studies of a 300-Mw(e) PuO/sub 2/-UO/sub 2/-fueled fast- breeder reactor are reported. Physics design parameters that effect fuel costs, full conservation, and reactor safety were evaluated for use in the selection of parameters for a reference design. The total breeding ratio varied from 1.1 to 1.5 in the range of parameters corsidered. Plutonium core loading ranged from 500 to 1500 kg. Doubling time was found to be reduced by high-density fuel and low steel content. A compromise figure on fuel-rod range of sizes (about 100 mils) yields a 5 operating reactivity and a small, negative sodium temperature coefficient. (J.R.D.)
Date:
November 15, 1959
Creator:
Greebler, P.; Aline, P. & Sueoka, J.
System:
The UNT Digital Library
Hot-Pressure Bonding of OMR Fuel Plates
Abstract: An alluminum-clad low-enrichment, uranium-alloy fuel element of flat plate configuration has been proposed for the Organic Moderated Reactor (OMR).
Date:
November 15, 1959
Creator:
Alm, G. V.; Binstock, M. H. & Garrett, E. E.
System:
The UNT Digital Library
PATHFINDER ATOMIC POWER PLANT COOLANT DISTRIBUTION TESTS. Final Report
Tests were made to determine the head loss coefficient through the inlet plenum of the Pathfinder reactor and to determine the now distribution among the fuel element nozzles for various operating conditions--with all three pumps operating at the same flow rate and with any combination of only two pumps operating at the same flow rate. A quarter-scale wooden model was used for the tests. Air was used as the fluld. The loss coefficient was determined to be 1.8 plus or minus 0.3. The velocities of flow through the fuel element nozzles were determined to be within plus or minus 5 per cent of average flow when all pumps are operating and within plus or minus 10 per cent of average flow when only two pumps are operating. (auth)
Date:
November 15, 1959
Creator:
Wilson, J. & Styles, R.
System:
The UNT Digital Library
PATHFINDER ATOMIC POWER PLANT FINAL REPORT MOISTURE DE-ENTRAINMENT TESTS IN TWO- AND FOUR-INCH DIAMETER TEST SECTIONS
Tests are described to determine the amount of entrained moisture occurring in the steam of the Pathfinder Power Plant as design parameters were varied. It was found that the height of the moisture separator and the velocity of the steam should be adjusted so that the permissible steam velocity for the particular height is not exceeded. If this is done, the amount of moisture entering the separator should be less than 1%, an amount of moisture that should not be difficult to remove with moisture separators. (J.R.D.)
Date:
November 15, 1959
Creator:
Wilson, J. & McDermott, M.
System:
The UNT Digital Library
Terminal Status Report for the Processing Refabrication Experiment
Introduction: A low-capacity, low-decontamination plant can be built, as part of a power reactor complex, to avoid long distance transfer of fuel to a high-capacity aqueous processing plant. Activation of such a complex, with the processing plant adjacent to the reactor it serves, could decrease the cost of the integrated fuel cycle. The study of this concept is a major objective of the Processing Refabrication Experiment (PRE).
Date:
November 15, 1959
Creator:
Sinizer, D. I.; Mattern, K. L. & Kendall, E. G.
System:
The UNT Digital Library
Carbide Fuel Development: Phase 1 Report, Period of May 15 to September 15, 1959
From foreword: This report describes the work accomplished including the Conceptual Design task and the initial results of the tasks on Facility Design and Fabrication, Fuel Fabrication and Evaluation, and Fuel Reprocessing.
Date:
October 15, 1959
Creator:
Bolomey, R.; Lazerus, S.; Sapir, J.; Sofer, G.; Steinmetz, H.; Strasser, A. et al.
System:
The UNT Digital Library
Collected Methods for Analysis of Sodium Metal
Methods for analyzing chemical impurities in sodium metal samples are presented. Chemical analysis was used to determine impurities in calcium, carbon, chromium, iron, lithium, nickel, oxygen, potassium, and zirconium. Spectrographic analysis was used to determine other impurities. Sodium samples obtained from experimental apparatus were analyzed by these methods. (auth)
Date:
October 15, 1959
Creator:
Perrine, H. E.
System:
The UNT Digital Library
Collected Methods for Analysis of Sodium Metal
Methods for analyzing chemical impurities in sodium metal samples are presented. Chemical analysis was used to determine the following impurities: Calcium, Carbon, Chromium, Iron, Lithium, Nickel, Oxygen, Potassium, and Zirconium. Spectrographic analysis was used to determine many other impurities. Sodium samples obtained from experimental apparatus operated as part of the work being conducted for Atomics International were analyzed by these methods.
Date:
October 15, 1959
Creator:
Perrine, H. E.
System:
The UNT Digital Library
Comments on Engineering Review of PRTR by Atomic Power Equipment Department
On a project of the magnitude and complexity of the Plutonium Recycle Test Reactor is was considered prudent engineering practice to obtain an independant design review by a competent off-site group that had not participated in any way in the development of the design. Accordingly, the Atomic Power Equipment Department of the General Electric Company was requested to make such an engineering review, and an Assistance to Hanford contract authorizing this work was approved by the Atomic Energy Commission in November, 1958.
Date:
October 15, 1959
Creator:
Reactor and Fuels Researcg and Development Operation Hanford Laboratories Operation
System:
The UNT Digital Library
Design Criteria for PRTR High Pressure Loop Project CAH-841
The PRTR pressure loop is an experimental facility to be installed in the Plutonium Recycle Test Reactor, Building 309, for use in testing the feasibility of reactor cooling systems utilizing high pressure water or steam as the coolant. It will furnish operating experience with such systems and serve to test fuel elements and components designed to operate in such environments.
Date:
October 15, 1959
Creator:
Fox, J. C.; Jakub, M. T. & Kelly, W. S.
System:
The UNT Digital Library