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THERMIONIC CONVERTERS FOR COMPACT NUCLEAR POWER PLANTS
A brief review of several thermionic nuclear power plants leads to the conclusion that the in-core concept is the most practical and useful system for space applications. Fundamental considerations indicate that emitter temperatures from 1500 to 1900 deg C are required for this concept, and that the high work function refractory metals are the best electrode materials for achieving the required performance and lifetime at these temperatures. The analytical and experimental technology developed for systematically defining the optimum materials and operatlng conditions is described, along with the significance of the experimental results obtained. These results have shown that the required performance can be obtained with the elementary cesium vapor diode converter. It was also shown that impurities cause pronounced effects on both performance and lifetime, and that their suppression or controlled utilization is an important aspect of the present approach. (auth)
Date:
August 1, 1962
Creator:
Rasor, N.S. & Weeks, C.C.
Object Type:
Report
System:
The UNT Digital Library
A STUDY OF LIQUIDS BY NEUTRON SCATTERING
A radial density-distribution function for liquids was derived with the aid of the assumption that vibrational modes exist in liquids. The distribution function was obtained only for elastically scattered neutrons. A study of si(s) curves as a function of temperature, as well as the energy distribution of the scattered neutrons, affords a way in which to construct a model describing the motion of atoms in a liquid. (auth)
Date:
March 1, 1962
Creator:
Persiani, P.J.
Object Type:
Report
System:
The UNT Digital Library
FEASIBILITY OF Pu$sup 239$-U$sup 235$-FUELED CORES TO PREDICT Pu$sup 239$- FUELED CORE DIMENSIONS
Use of Pu/sup 239/ -- U/sup 235/-fueled fast critical assemblies to estimate properties of Pu/sup 239/-fueled assemblies is of interest because of safety considerations and limited plutonium availability. Bare and reflected homogeneous cores and reflected two-region cores are considered. The fuel, 5% by volume, is assumed to be Pu/sup 239/ and U/sup 235/ of various fuel composition ratios for the homogeneous cores. For the tworegion cores the 5% fuel volume is Pu/sup 235/ in the central region and U/sup 235/ in the outer core region. Core diluents, simulating fertile, structural, and coolant materials, are assumed identical in all cases. it is estimated that construction of the reflected two- region core with ratio of central core region volume to total core volume of 0.1 will theoretically decrease the calculated error in prediction of the critical size of a corresponding solely Pu/sup 239/-fueled assembly by a factor of about 10 to 20. (auth)
Date:
June 1, 1962
Creator:
Meneghetti, D. & Ishikawa, H.
Object Type:
Report
System:
The UNT Digital Library
SOME CALCULATIONS PERTAINING TO FAST REACTOR SAFETY
A program on fast reactor safety is described in which the theoretical aspects, techniques and accuracy of calculation, and associated safety predictions are considered. One topic deals with the success obtained in analyzing complicated meltdown configurations using two-dimensional, twogroup diffusion theory. Another topic deals with the dependence of energy yield in a fast reactor explosion on the shape of the power distribution in the core. The results of a brief survey are included dealing with the change of dynamic reactor behavior during various startup accidents as a consequence of a reduction in the delayed-neutron fraction. (J.R.D.)
Date:
February 1, 1962
Creator:
O'Shea, D.M.; Okrent, D. & Chaumont, J.M.
Object Type:
Report
System:
The UNT Digital Library
An Out-of-Pile Electron Irradiation Circulating Loop for Fouling Studies
An out-of-pile electron irradiation circulating loop was developed to study the problem of fuel element fouling in organic cooled reactors. Unlike pyrolytic loops and capsules, the irradiation loop operates at temperatures, pressures, fluid dynamics, geometry, and radiation damage which closely resemble reactor conditions. The loop is run in conjunction with the 4-kw -- 6-Mev electron linear accelerator. The thickness of the fouling film deposited in the irradiation cell is measured and the film subjected to chemical analysis. Test results showed that clean coolant did not foul; reactor coolant (Core II) with particulate matter fouled; heaviest film was found in the radiation area; and lower fouling occurred at high velocity. (auth)
Date:
August 1, 1962
Creator:
Mengelkamp, R. A.; Hudson, P. S. & Hillyer, J. C.
Object Type:
Report
System:
The UNT Digital Library
NUCLEAR INSTRUMENTATION FOR SCINTILLATION AND SEMICONDUCTOR SPECTROSCOPY
A manual is presented for those who use or service the transistorized instruments for nuclear spectroscopy: the transistor amplifier; the snip-snap single-channel analyzer; the fast coincidence unit; and the biased amplifier and linear gate. A general description is given for each instrument along with the specifications, a description of the circuit, and a procedure for initial testing. (auth)
Date:
May 1, 1962
Creator:
Emmer, T.L.
Object Type:
Report
System:
The UNT Digital Library
CHEMICAL PROCESSING OF COATED PARTICLE FUELS
BS>Laboratory studies on the processing of graphite-base fuel elements containing pyrolytic carbon- or Al/sub 2/O/sub 3/-coated particles are reviewed. Potential processes for recovering U and Th from irradiated elements include grinding followed by acid leaching, and, burning and subsequent dissolution of the oxide ash. Disintegration in 90% HNO/sub 3/ was briefly evaluated as a method for determining the integrity of coated particles dispersed in graphite matrices. (auth)
Date:
April 1, 1962
Creator:
Ferris, L.M.
Object Type:
Report
System:
The UNT Digital Library
ULTRA HIGH TEMPERATURE REACTOR EXPERIMENT (UHTREX) HAZARD REPORT
UHTREX utilizes a high-temperature, He-cooled, graphite moderated reactor employing unclad, refractory fuel elements. The reactor is designed to produce a msximum thermal power of 3 Mw and a maximum exit He temperature of 2400 deg F. The purpose of the experimert is to evaluate the advantages of the simple fuel against the disadvantages of the associated operation of a contaminated coolant loop. The mechanical and nuclear design of the reactor and related apparatus are described, discussed, and evaluated from the standpoint of hazards associated with conduct of the experiment. The building design and characteristics of the site are also examined from the same standpoint. The probable effects of operational errors and component failures are studied. The conseqnences of credible accidents are not considered to be catastrophic for either operating personnel or personnel in surrounding areas. (auth)
Date:
March 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
THE ELECTRIC-DIPOLE GAMMA-RAY STRENGTH FUNCTION FOR HEAVY EVEN-EVEN NUCLEI
None
Date:
August 1, 1962
Creator:
Carpenter, R.T.
Object Type:
Report
System:
The UNT Digital Library
High Resolution Proton Magnetic Resonance Spectra of Silylphosphine, Silylarsine, Germylphosphine and Germylarsine
The proton magnetic resonance spectra of SiH/sub 3/PH/sub 2/, SiH/sub 3/ AsH/sub 2/, GeH/sub 3/PH/sub 2/, and GeH/sub 3/AsH/sub 2/ are describe d and interpreted. (auth)
Date:
August 1, 1962
Creator:
Drake, J. E. & Jolly, W. L.
Object Type:
Report
System:
The UNT Digital Library
ELECTROLYTIC DISSOLUTION OF NUCLEAR FUELS. PART III. STAINLESS STEEL (304) IN NITRATE SOLUTIONS
The potential-current density relationships for 304 stainless steel dissolution in a nitrate system were studied as a function of solution composition and temperature in order to optimize the conditions for electrolytic dissolution of ihis material. In the nitrate system, the anodic dissolution of steel takes place in the transpassive region. Under some conditions, deviations from Tafel behavior are observed which depend greatly on the nitrate and hydrogen ion concentration, and on temperature. A discussion of passivity, transpassivity, secondary passivation, the limiting current density, and the effect of alloy composition on the dissolution behavior is given. It was found that at temperatures above 60 deg C efficient dissolver operation should be possible over a wide range of solution compositions and at current densities up to 2 amp/cm/sup 2/. (auth)
Date:
June 1, 1962
Creator:
Aylward, J. R. & Whitener, E. M.
Object Type:
Report
System:
The UNT Digital Library
SPECTRAL SHIFT CONTROL REACTOR BASIC PHYSICS PROGRAM-THEORETICAL ANALYSIS. PART I. ANALYSIS OF EXPERIMENTS
The data obtained in the series of SSCR experiments and the theoretical analysis of the results demonstrate the spectral shift control principle clearly. The critical assemblies show that considerably larger amounts of fuel can be loaded into a core moderated with a mixture of D/sub 2/O and H/sub 2/O than in one moderated with H/sub 2/O alone. The measured Cd ratios and the calculated neutron balances show that the excess reactivity associated with this extra mass is held up in the resonance absorption of the fertile material. This leads to a higher conversion ratio system than the same sized lattice controlled by thermal neutron poison. Very good agreement is obtained between experimental and theoretical values for nearly all important parameters over the entire range of D/ sub 2/O concentration in the moderator. This agreement confirms the use of the BPG-four-group treatment, and demonstrates the calculation of reactor criticality and other measured integral parameters from fundamental cross sections without the use of intermediate fitting functions. The reactor leakage part of the calculational model is shown to be accurate by the good agreement obtained between calculated and experimental values of age, criticality, change of criticality with moderator height, and the comparison between …
Date:
March 1, 1962
Creator:
Wehmeyer, D.B.; Doederlein, J.M.; Roach, K.E. & Wittkopf, W.A.
Object Type:
Report
System:
The UNT Digital Library
Chemical Technology Division, Unit Operations Section Monthly Progress Report, August 1961
Engineering studies of a 6-in.-ID countercurrent foam column were started. The capacity of Mark II Stacked Clone Contactor was found to be limited by gradually increasing entrainment. Thoria sols were prepared from trough denitrator products in a conical bottom tank agitated by circulation through an external centrifugal pump. Experimental results agreed favorably with predicted results for the reaction of CO with CuO and fur the simultaneous reactions of H/ sub 2/ and CO with CuO. In a 2-in. dissolver, Modified Zirflex dissolution rates of Zircaloy-2 increased non-linearly with F/S. The 250-ton prototype shear was received and installed. Priorities were established for development work on graphite fuels. A total of 5 SR Core I fuel clusters were mechanically dejacketed from the second carrier shipment of SRE Core I fuel. Degradation of NaK to a solid or waxy form required that the slugs be ejected by the jackscrew. The NaK disposal system operated satisfactorily for the Iast ~3000 ml of NaK. A procedure for estimating the temperature rise of spent reactor fuels during shipping gives reasonably good results. A puIse column was operated with the light phase c nt-nuous using an external phase separator. Zirconium ates were dissolved by HF in molten …
Date:
March 1, 1962
Creator:
Whatley, M. E.; Haas, P. A.; Horton, R. W.; Ryon, A. D.; Suddath, J. C. & Watson, C. D.
Object Type:
Report
System:
The UNT Digital Library
Annual Progress Report on Fuel Element Development for Fiscal Year 1961
Progress in fuels and materials development is summarized. Major areas of investigation include a materials study by means of sample fuel plates containing uranium alloys or cermets, burnable poisons, non-uniform fuel and poison distributions and clad with various aluminum alloys; and an engineering study of fuel element geometries optimized in heat transfer, hydraulics, and materials strength. Up to 45 wt% U-Al alloys, 6 to 65 wt% UO/-Al and U3O6-Al dispersions, including enrichments ranging from 20% to 93%, were tested to 70% burnup in de-ionized water at 200 deg F in the MTR. Their performance at higher temperature is still being investigated. Test results for the MTR conditions indicate that all of the compositions investigated to date will successfully withstand even the longest irradiation at these conditions if properly fabricated. Some high strength aluminum alloy claddings, not yet fully tested, show some peculiar surface effects which may be related to corrosion. Metallographic studies of irradiated cermets reveal a reaction'' (diffusion) zone produced around UO/sub 2/ particles in contact with aluminum. These zones are being studied by means of x-ray diffraction, electron microscopy, and electron microprobe analysis. From engineering studies has come promise of improved heat removal and lower pumping requlrements for …
Date:
March 1, 1962
Creator:
Gibson, G. W. & Shupe, O. K.
Object Type:
Report
System:
The UNT Digital Library
NEUTRON INTERACTIONS IN LIQUID PARA- AND ORTHO-HYDROGEN
An experimental method is described which utilizes a simple Fermi-type chopper and conventional time-of-flight techniques in conjunction with an electron linear-accelerator pulsed neutron source to study inelastic scattering of thermal neutrons from liquid para- and ortho-para-hydrogen. The overall general agreement between theory and experiment supports the concept that the molecules in liquid hydrogen behave nearly as they would in a perfect gas. For liquid ortho-para-hydrogen, however, the extra broadening of the energy distribution of scattered neutrons suggests that some effects of the liquid state may be present. Similar observations on para-hydrogen indicate that, in this liquid as well, the motion of the molecules is affected by weak intermolecular forces. (auth)
Date:
August 1, 1962
Creator:
Whittemore, W L & Danner, H R
Object Type:
Report
System:
The UNT Digital Library
ECONOMIC FACTORS OF MFP THERMOELECTRIC GENERATORS. Interim Report
Mixed Fission Products (MFP) for use as a heat source for thermoelectric generators will become increasingly available in the coming years. The Atomic Energy Conamission sponsored program on solidification of nuclear wastes is now entering the hot-bench scale test phase. During this phase approximately 5000 thermal watts of two year old MFP could be produced monthly. Two different types of hot calcination pilot plants are planned for installation at the Hanford National Laboratories in the 1964 to 1966 time period. Each of these plants should be able to produce 160,000 thermal watts of two year MFP and 16,000 thermal watts of ten year MFP on a monthly basis. During this phase, MFP costs should be less than 15 per ihermal watt for two year MFP and 50 for ten year MFP. This cost includes operation of the plant solely to obtain heat sources and sealing the MFP into fuel containers. A full scale plant for a 15,000 Mw(e) nuclear economy is estimated to produce four to five times as much MFP as either of the pilot plants. Costs will be dependent upon AEC policy in effect at the time the plant is operating. lf the policy indicates that the full …
Date:
June 1, 1962
Creator:
Barmat, M.
Object Type:
Report
System:
The UNT Digital Library
DEVELOPMENT OF PLUTONIUM-BEARING FUEL MATERIALS. Monthly Progress Letter for Month of January 1962
Development work leading to comparison of co-precipitated and mechanically blended UO/sub 2/-- PuO/sub 2? powders is reported in which special emphasis was placed on blending trials, pellet sintering studies, and subsequent evaluation of pellets made with blended material. Homogeneity studies indicate that currently used procedures are unsatisfactory because particle buildup occurs during blending. Powder preparation via the oxalate route was continued along with PuO/sub 2/ moisture pickup studies. Homogeneous precipitation studies on UO/ sub 2/ were continued to determine feasibility of direct preparation of dense UO/ sub 2/-- PuO/sub 2/ feed materials. Plasma-torch-produced PuO/sub 2/ spheres are being evaluated. (J.R.D.)
Date:
February 1, 1962
Creator:
Puechl, K.H.
Object Type:
Report
System:
The UNT Digital Library
SNAP Programs. M-1 Monte Carlo Radioisotope Shielding Code. Final Report
The M-1 code is a Monte Carlo code that applies to cylindrical geometry when solving for the flux from a pre specified radiation source. The source is a gamma and beta emitter and the solution is for the flux of each energy group and of each region of interest in regard to the emitter. A region is a volume of the system bounded by two planes perpendicular to the axis of symmetry and two cylinders (one cylinder if the region includes the axis of symmetry). The code can be used to solve for a maximum of 30 energy groups and 280 regions. The M-1 is coded in Fortran for a 32,000-word 7090 and requires that the energy intervals be prespecified as well as a complete description of the geometry of the system. A specification of materials in the system must also be given. The number of particles to be followed must be specified by the user. Since the technique of splitting can be employed here and so that splitting can occur, a description of the manner in which the system is divided (geometrically) must also be given by the user. A detailed description of the input required by the code …
Date:
May 1, 1962
Creator:
Kniedler, M. J.
Object Type:
Report
System:
The UNT Digital Library
DEVELOPMENT OF PLUTONIUM-BEARING FUEL MATERIALS. Monthly Progress Letter for Month of February 1962
Major effort during the report period was directed toward the in-pile testing program. Component PuO/sub 2/ and UO/sub 2/ powder blending trials were continued along with coprecipitated powder characterization studies. Arrangements for irradiation of short exposure rabbit capsules were made. (J.R.D.)
Date:
March 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
SUBJECT HEADINGS USED BY THE U.S.A.E.C. DIVISION OF TECHNICAL INFORMATION
A list is presented of subject headings used by the U.S.A.E.C. Division of Technical Information for indexing literature in the field of nuclear science and technology. (C.H.)
Date:
August 1, 1962
Creator:
Hargrave, C.W.
Object Type:
Report
System:
The UNT Digital Library
THE CHEMISTRY OF NIOBIUM IN PROCESSING OF NUCLEAR FUELS
The chemical and industrial literature and laboratory work concerned with processing of niobium-containing nuclear fuels are reviewed. 57 references. (auth)
Date:
February 1, 1962
Creator:
Gens, T.A.
Object Type:
Report
System:
The UNT Digital Library
Selection of Core Design No. 1 for Type 5 Replacement Cores in SM-1 and SM-1A
Nuclear and thermal analyses were performed to determine the characteristics of the Type 5 core in the SM-1 and SM-1A reactor plants as a function of geometry and composition. The following nuclear properties were investigated: core energy release, maximum midlife reactivity, average fuel burnup fraction, B-10 reactivity coefficient, and power distribution. Thermal parameter surveys determined the effects of channel thickness and power distribution upon the DNBR, nominal and hot channel thermal performance, and fuel plate thermal stress. From the nuclear and thermal analyses, a Type 5 core reference design was selected with fuel plates of 70-mil plate thick ness, 7-mil clad thickness, and 38 wt % UO/sub 2/ in the matrix, having initial core loading o4 108 Kg U/syup 235 and 260 gm B/sup 10/. (auth)
Date:
July 1, 1962
Creator:
Davidson, S. L. & Paluszkiewicz, S.
Object Type:
Report
System:
The UNT Digital Library
CHEMICAL ENGINEERING DIVISION SUMMARY REPORT, APRIL-JUNE 1962
Research and development progress is reported on chemical-metallurgical processing, fuel cycle applications of voiatility and fluidization techniques, calorimetry, reactor safety, energy conversion, determination of nuclear constants, and routine operations. (M.C.G.)
Date:
August 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
MODIFICATION OF THE EXPERIMENTAL BOILING WATER REACTOR (EBWR) FOR HIGHER- POWER OPERATION
Supplement to ANL-5607. Alterations and additions made to the Experimental Boiling Water Reactor (EBWR) plant to permit operation at power levels up to I00 Mw(t) are described. Topics covered include over-all system modifications and additions, nuclear component modifications and additions, and reboiler plant component description. (M.C.G.)
Date:
April 1, 1962
Creator:
Matousek, J.F. comp.
Object Type:
Report
System:
The UNT Digital Library