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A suggested future Spade and Snoopy program for Pluto effort

Description: This memorandum elaborates on items discussed in a meeting held July 13, 1961 a suggested Spade and Snoopy Program for the Pluto effort. Topics were specific Tory II-C features, basic studies, and miscellaneous items.
Date: July 13, 1961
Creator: Goldberg, E.
Partner: UNT Libraries Government Documents Department

Design Study of Portable Thermoelectric Nuclear Systems

Description: Design studies were performed and costs were estimated for an air transportable, 10 Mw(t), pressurized light water thermal circulation reactor, combined with a direct conversion thermoelectric generator and static electrical inversion equipment. This TCR-TE'' concept appears to have potential for ultimate use as a remote unmanned power station. Based on an extrapolation of present reactor technology and on assumed thermoelectric materials properties forecasted to January 1, 1963, a net a-c electrical output of 315 Kw is estimated, assuming the use of 80 deg F local water for cooling purposes. An alternate concept using 80 deg F air for cooling produces 271 Kw, net. These electrical output figures can be improved significantly through a recommended research and development effort. The design and construction of a prototype plant is also recommended. (auth)
Date: July 1, 1961
Creator: Chajson, L.; DelCampo, A. R. & Kellogg, H.B. et al
Partner: UNT Libraries Government Documents Department

Thermoelectric Nuclear Fuel Element Quarterly Progress Report, April-June 1961

Description: Uranium-bearing thermoelectric compounds are now being prepared by tantalum bomb melting and by the hydride process. Tests of devices made up from these compounds indicate that the main fabrication problems are densification and contact bonding. Data from a hot-swaged pellet and a swaged device of US/sub 2/ indicate some promise for that compound. Improvements in techniques of thermoelectric parameter measurements include programming of automatic test data recording at desired intervals around the clock; increased accuracy and versatility of measurements through use of a newly-constructed adjustable precision resistor; and a method for measuring which should lead to an experimental means for determining the thermoelectric figure of merit, Z. Potential profile studies on PbTe pelleta are yielding important information on contact resistance parameters. A fission-fired thermoelectric generator is being prepared for the next in-pile test. (auth)
Date: July 10, 1961
Partner: UNT Libraries Government Documents Department

Fabrication of Fuel Rods by Tandem Rolling

Description: From introduction: "The purpose of this report is to present the details of the exploratory and developmental work on tandem rolling, and the subsequent fabrication of tandem rolled fuel rods for irradiation testing in the Vallecitos Boiling Water Reactor."
Date: July 1961
Creator: Lingafelter, J. W.
Partner: UNT Libraries Government Documents Department

The RBU Reactor-Burnup Code: Formulation and Operation Procedures

Description: Report discussing the computer program RBU, which calculates the neutron, reactivity, and isotopic history of a nuclear reactor in such a way as to facilitate the predictions of fuel costs and reactor performance. This report documents RBU's various calculations and operating procedures.
Date: July 1961
Creator: Triplett, J. R.; Merrill, E. T. & Burr, J. R.
Partner: UNT Libraries Government Documents Department

Health and Safety Laboratory Fallout Program Quarterly Summary Report: March 1, 1961 - June 1, 1961

Description: Report that summarizes multiple laboratories' reports on global fallout deposition. Reports include data on Strontium-90 deposition recorded by the Health and Safety Laboratory, data from other laboratories, related interpretive reports, and recent publications related to fallout.
Date: July 1, 1961
Creator: Hardy, Edward P., Jr.; Rivera, Joseph & Frankel, Robert
Partner: UNT Libraries Government Documents Department

Precipitation of Neptunium Peroxide

Description: Optimum conditions were determined for the precipitation of neptunium(IV) peroxide from nitric acid solutions. The results indicate that the precipitation could be applied successfully on a plant scale. Data are presented for the solubility of neptunium peroxide in solutions of nitric acid and hydrogen peroxide. The solubility is less than 10/sup -4/M in 1.5 to 2.5M nitric acid containing 4.5M hydrogen peroxide. Two crystalline modifications of the peroxide were prepared; these two crystalline structures were similar to structures previously reported to plutonium peroxide. (auth)
Date: July 1, 1961
Creator: Burney, G. A. & Dukes, E. K.
Partner: UNT Libraries Government Documents Department

DETERMINATION OF LIFE-TIME OF THE B$sub 4$C-IN-TUBES CONTROL ROD

Description: The lifetime of a B/sub 4/C-in-tubes'' control rod may be limited by either nuclear worth depreciation or internal gas pressure buildup as helium is formed and released from the B/sub 4/C particles. Nuclear life is enhanced by thin tube walls for any given O.D. of tubing and by high B/sub 4/C density, but the time to failure from gas pressure buildup is increased by the opposite. The optimum B/sub 4/C density and tube wall thickness were calculated for maximum control rod life, using several release fractions of generated helium and a number of allowable cladding stress values. It was estimated that lifetimes of l0 or more years for B/sub 4/C-in-tubes control systems are achievable through the proper selection of tubing material and B/sub 4/C density values. The B/sub 4/C densities required for maximum lifetimes are in a range that can be obtained by vibratory compaction followed by swaging. (auth)
Date: July 19, 1961
Creator: Megerth, F.H.
Partner: UNT Libraries Government Documents Department

THE SNAP II POWER CONVERSION SYSTEM TOPICAL REPORT NO. 12. BOILER DEVELOPMENT

Description: The SNAP II boilers which were designed are summarized. As shown by test results from the three boilers which were tested, a continuous progress in design was achieved. These designs were based on test data from both the SNAP I and SNAP II programs. As the quantity of data increased, physical models describing the heat transfer process were developed. These physical models provide the necessary correlation parameters which permit the extension of existing data to advanced design. Preliminary test sections were designed on the assumption that an allvapor nmodel which ignores the presence of the liquid phase during forced convection boiling could be used to describe the process quantitatively. The conventional Dittus-Boelter equation was applied with the increase in the vapor flow along the tube being ascribed to liquid evaporation. The assumption led to a design that fell short by about an order of magnitude since the exit qualities were only in the range of 10%, far less than required for complete vaporization. As a result, a revision in the concept of the mechanics of boiling was found necessary and a theoretical analysis was formulated, based on a dry wall'' or dropwise'' type boiling phenomenon. The test results of the preliminary test sections and the SNAP I boiler were plotted on the basis of dry-wall boiling parameters containing the area mean temperature difference and mass velocity. A conservative design curve was established and used to design the thirteen tube boiler. The design was found by test to be conservative, and the measured performance and the degree of conservatism were found to be within the expected spread in earlier test data. Dropwise boiling pictures the heat transfer as occurring directly from the wall to the drop through a film created by the vapor being ejected from the underside of the drop. ...
Date: July 17, 1961
Creator: Gido, R.G.; Koestel, A.; Haller, H.C.; Huber, D.D. & Deibel, D.L.
Partner: UNT Libraries Government Documents Department

THE EXPERIMENTAL BERYLLIUM OXIDE REACTOR. MARITIME GAS-COOLED REACTOR PROGRAM

Description: LIUM OXIDE REACTOR. MARITIME GAS-COOLED The Experimental Beryllium Oxide Reactor, EBOR, will be constructed at the National Reactor Testing Station as the AEC portion of the joint Maritime Administration--AEC Maritime Gas Cooled Reactor Program. The ultimate goal of the Program is the development of nuclear power plants employing a helium cooled and beryllium oxide moderated reactor directly coupled to a closed cycle gas turbine. The objective is to obtain compact nuclear engines suitable for use either in a merchant ship propulsion system or an intermediate size central station power plant in the 20 to 100 Mw(e) size range. The EBOR is a l0 Mw(t) test of the basic fuel element and moderator designs. It is capable of being up-graded in power at a later date to a test of the nuclear reactor turbine concept. The objective of the experiment is outlined. The principal reactor components to be tested and the test facility are described. (auth)
Date: July 1, 1961
Creator: Moore, W.C.
Partner: UNT Libraries Government Documents Department

USE TEST COMPARISON OF TBP DILUENTS

Description: Several diluents for possible use in TBP Purex Plant solvent were tested. The tests included nitric --nitrous acid degradation, fission prcduct distribution under simulated plant conditions, emulsillcation, and radiolysis. The order of quality of four diluents is n-dcdecane> Soltrol 170> Shell Code 85030(82000)> Shell E-2342. (D.L.C.)
Date: July 1, 1961
Creator: Mendel, J.E.
Partner: UNT Libraries Government Documents Department

OXIDATION OF GRAPHITE UNDER HIGH TEMPERATURE REACTOR CONDITIONS

Description: A kinetic study was conducted to provide information on oxidation of reactor graphites in the temperature range of 450 to 675 deg C and on the effects of reactor environment on oxidation rates. Among the parameters studied were chemical reactivity of the graphite, prior oxidation, a high intensity gamma flux during oxidation, variation of the surface-to-volume ratio of the graphite specimens, neutron bombardment prior to oxidation exposure, and gas flow rates. Rate equations showed apparent activation energies of 50 kcal/mole in the absence of radiation and 30 kcal/mole in the presence of a 1 x 10/sup 6/ r/hr gamma flux. (auth)
Date: July 1, 1961
Creator: Dahl, R.E.
Partner: UNT Libraries Government Documents Department

EXPERIMENTAL MEASUREMENTS OF THE SUCTION HEAD REQUIRED BY THE HALLAM PROTOTYPE FREE SURFACE SODIUM PUMP

Description: Hydraulic tests were made on the Hallam Prototype Free-Surface Sodium Pump to determine the net positive suction head (NPSH) required at various sodium flow rates. Pump performance data were also collected. The results indicate that an NPSH of 22 ft sodium is required at the design flow rate of 7200 gpm at approximates 1000 deg F, agreeing with computed values, and that the pump is designed with a safety margin of slightly over l0%. (D.L.C.)
Date: July 25, 1961
Creator: Atz, R.W.
Partner: UNT Libraries Government Documents Department

PLASTIC STRAIN IN THIN FUEL ELEMENT CLADDING DUE TO UO$sub 2$ THERMAL EXPANSION

Description: Results are presented of short time, high heat flux capsule irradiations to determine the magnitude of plastic strain induced in the capsule cladding due to UO/sub 2/ thermal expansion. Annular geometry capsules were employed with both internal and external cladding and forced convection cooling of both surfaces to simulate the reference superheat fuel element design. Both 304 stainless steel and 3003 aluminum claddings of various thickness were used to determine the influence of cladding strength. The initial gap spacing was purposely varied as an additional parameter. The program was undertaken as a part of the effort to evaluate plastic strain cycling as life limitation in the use of thin fuel element cladding. An over-all summary of the experimental results is included. (auth)
Date: July 1, 1961
Creator: Lyons, M.F.; Comprelli, F.A.; Hazel, V.E. & Townsend, H.E.
Partner: UNT Libraries Government Documents Department

EXPERIMENTAL INVESTIGATION OF THE EFFECTS OF ULTRASONIC VIBRATION ON BURNOUT HEAT FLUX WITH BOILING WATER. Final Summary Report, October 3, 1960-July 31, 1961

Description: Experimental results were obtained on the effect of an ultrasonic field on the burnout heat flux for water flowing at atmospheric pressure, through an annular flow channel formed by a 1/4-in.-diameter electrically heated tube and a concentric glass tube of 3/4-in. ID. The active length of the central heating element was 5 1/2 in. The ultrasonic transducer, which was operated at 25,000 cps and a maximum electrical input of 300 watts, was located at the inlet end of the flow channel. The ultrasonic waves were propagated in the water in the direction of flow and thus parallel to the surface of the heating element. Burnout conditions covered channel inlet flows from 1.61 to 6.25 ft/sec and subcooling from 16 to 28 deg F. No effect of the ultrasonic field on the burnout heat flux or on the visible boiling phenomena at burnout conditions was detectable. During boiling at heat fluxes well below burnout, the effect of the ultrasonic field was a reduction in the diameter of the envelope of bubble activity surrounding the heating element. Visual inspectibn appeared to show that this reduction was associated with a smaller average bubble size and a greater frequency of bubble formation. However, all evidence of the presence of the ultrasonic field vanished as the flow velocity increased or as the heat flux increased to the burnout level. (auth)
Date: July 31, 1961
Creator: Romie, F.E. & Aronson, C.A.
Partner: UNT Libraries Government Documents Department

Dynamic Properties of Heterogeneous Water Reactors

Description: The types of tests performed in SPERT-I, and the tests proposed for SPERT-II and -III, are described. These reactors are described, and factors influencing their dynamic behavior are discussed. The tests are classed as static, step, ramp, and oscillatory. The correlation between the test results and the reactor dynamic safety characteristics (stability, self-shutdown under excursion conditions, etc.) is investigated. (T.F.H.)
Date: July 20, 1961
Creator: Forbes, S. G. & Nyer, W. E.
Partner: UNT Libraries Government Documents Department

SPERT IV HAZARDS SUMMARY REPORT

Description: Spert IV is a large pool-type experimental facility for reactor kinetic studies. These studies will include power excursion and instability tests for a variety of reactor designs. Since the Spert IV experimental program requires the performance of tests which will approach, and may exceed the threshold of reactor destruction, the probability of occurrence of the maximum possible accident is not negligible compared with that of other possible accidents. The maximum possible accident for this facility is considered to be a severe nuclear excursion which results in the destruction of the reactor building and the release of 100% of the accumulated fission product inventory of the atmosphere in a steam cloud. The fission product source assumed in the analysis of this accident is an upper limit in view of the nature of the tests to be performed and the heat removal capacity of the system. This postulated accident is independent of the details of core and control system design and is valid for all cores anticipated for use in the experimental program. The major hazards present in the operation of this facility, the precautions to be taken to reduce the probability of an accident, and the consequences of the maximum possible accident are discussed. It is concluded that the proposed method of operation will minimize the hazard to operating personnel, and that the site location will make possible the operation of the Spent IV facility without hazard to the general public. (auth)
Date: July 1, 1961
Creator: Bentzen, F. L. & Crocker, J. G.
Partner: UNT Libraries Government Documents Department

THERMODYNAMICS AND STATISTICAL MECHANICS OF A THREE-LEVEL MASER

Description: The population distribution and the associated thermodynamic properties of a three-level maser are calculated. By treating the three spin states as individual chemical species, it is possible to write the partition function, free energy, internal energy, entropy, and chemical potential for each species. The principle of minimum entropy production is used to derive an equation of reaction equilibrium. This equation is used in separate thermodynamic and statistical mechanical calculations to obtain the normalized population distribution. This result agrees in first order with that obtained by means of the rate equations. Competitive relaxation processes and partial saturation effects are discussed. The saturation parameter used is the one defined in the treatment of the Overhauser effect. An explicit expression is obtained for this parameter in terms of the various transition probabilities, and its effect on maser performance is discussed. (auth)
Date: July 1, 1961
Creator: Barker, W.A.
Partner: UNT Libraries Government Documents Department

STRUCTURES AND PROPERTIES OF URANIUM-FISSIUM ALLOYS. Final Report- Metallurgy Program 4.1.23

Description: A study was made of the phase relations and the properties of uranium-- fissium alloys which have compositions bracketing that intpnded for the first core loading of Experimental Breeder Reactor II. The fissium aggregate in the alloys consisted of the elements Zr, Nb, Mo, Ru, Rh, and Pd. Phase relations are shown to parallel closely those in the dominant U--Mo--Ru ternary system. The uranium gamma phase is stabilized down to 552 deg C, while the beta phase is entirely suppressed at high fissium contents. Certain crystallographic data are given and the minor phases that occur in the alloys are identified. In cast and gammaquenched alloys the retention of the high-temperature gamma phase produced low hardness and low density. The thermal expsnsion behavior of the alloys is shown to be dependent upon composition and prior thermal history. Thermal conductivity data are presented for uranium and the uranium-- fission alloys. The thermal conductivities of the alloys decrease with increasing fissium concentration. (auth)
Date: July 1, 1961
Creator: Zegler, S.T. & Nevitt, M.V.
Partner: UNT Libraries Government Documents Department

THE PHYSICS DESIGN OF THE EBR-II

Description: The physics design problems of the EBR-II are summarized. These include analysis of the EBR-II engineering design as well as applicable zero-power critical experiments. Pertinent reactor safety problems are reviewed. Safety considerations bearing on normal plant operation and manipulations within the reactor are emphasized. The implication of controlled in-pile meltdown experiments is considered. Irradiation damage and metallurgical phase phenomena are summarized and related to reactivity. The nuclear performance of the system is considered in terms of actual plant operation. The predicted shift of both power and reactivity from core to radial reflector is described. (auth)
Date: July 1, 1961
Creator: Loewenstein, W.B.
Partner: UNT Libraries Government Documents Department

POWER-TO-VOID TRANSFER FUNCTIONS

Description: Variations in the distribution of steam bubble, the "void" distribution, in a boiling channel as a function of changes in heating power were studied. A rectangular test tube, of 1.11 x 4.44-cm cross section and 127-cm height, was inserted in a forced-circulation pressure loop. The tube was heated by passing an a-c current through the tube walls. A power oscillator was built which could give a 10% peak-topeak sinusoidal power modulation at any frequency in the interval from 0.01 to 10 cps. Variations in the volume fraction of steam were observed by means of a gamma densitometer built for the purpose. Accurate void profiles could be taken by traversing the test channel vertically and horizontally. With the void detector stationary at a given height, the amplitude and phase delay of the steam void variations were measured in the frequency range mentioned. The signal from the gamma detector was passed to a harmonic analyzer built for the experiment. This instrument could pick out the void variations coherent with the power variation in the presence of much greater random signal variations caused by the boiling process. The frequency response of steam void was measured at 4 different pressures ranging from 27.2 to 68 atms, at conditions comparable to those in pressurized boiling water reactors. Void phase and void amplitude are plotted as functions of frequency, and the data are also presented in tables. The most important result of the experiments is to show that the void response falls off at a frequency that is much lower than that predicted by theoretically derived power-to-void transfer functions used previously in reactor calculations. Also, the void amplitude in the lower part of the channel was larger than expected. By taking into account the pressure changes in the channel caused by the power variations, an expression ...
Date: July 1, 1961
Creator: Christensen, H.
Partner: UNT Libraries Government Documents Department