WorldCat Identities

Phillips Petroleum Company Atomic Energy Division

Works: 714 works in 1,410 publications in 1 language and 8,270 library holdings
Genres: Periodicals  Handbooks and manuals 
Classifications: QC373.S7, 539.775
Publication Timeline
Most widely held works by Phillips Petroleum Company
Scintillation spectrometry : gamma-ray spectrum catalogue by R. L Heath( Book )

5 editions published between 1957 and 1964 in English and held by 136 WorldCat member libraries worldwide

Chemical processing technology quarterly progress report( )

in English and held by 21 WorldCat member libraries worldwide

Organic coolant reactor program quarterly report : January 1-March 31, 1961 by J. R Huffman( Book )

21 editions published between 1960 and 1964 in English and held by 21 WorldCat member libraries worldwide

Experimental and calcuiated results show that flux flattening is accomplished by moving the boron burnable poison in ETR fuel elements from the fuel region to the sideplates. MTR shim rod calibrations were made for Cycle 146- B by distributed poison techniques. Excess reactivities calculated from these calibrations by three definitions give maximum values of from approximately 13.5 to 15.5%. The Advanced Reactivity Measurement Facility (ARMF) is now supplementing the Reactivity Measurement Facility (RMF) in making reactor physics measurements on small samples. The design aim of the ARMF is to achieve the maximum sensitivity and reproducibility possible with more stable operation and better control than in the RMF. Heat transfer calculations are being made on a thorium slug in the form of a thick-walled tube, which it is proposed to irrddiate, instead of the present solid slug, for U/sup 2//sup 3//sup 3/ production in the MTR high flux. Experimenthl techniques are being developed for measuring the l/E component of neutron flux in several experimental positions of the RMF. Irradiation of a highly enriched 54 wt.% UO/sub 2/--Al fuel plate sample to 56% U/sup 2//sup 3//sup 5/ burnup produced a reaction (flssion damage) zone around the UO/sup 2/ particles. Hydraulic tests made on roughened fuel plates indicated roughening as a possible means of increasing the heat transfer rate during forced convection cooling. Analyses of the total cross sections for Pu/sup 2//sup 4//sup 1/ and Pa/sup 2//sup 3//sup 1/ were ess entially completed, and a final report for Pu/sup 2//sup 4//sup 1/ was submitted for publication. Measurements of U/sup 2//sup 3/ eta values using the Mn bath technique with the monochromatic crystal spectrometer neutrons were continued. Experimental results to date are primarily concerned with measurements of the corrections required. Scattering measurements for samples of U/sup 2//sup 3//sup 3/ were undertsken on the fast chopper. Various integral cross sections in reactor spectra were undertaken. The 27,000 b capture cross section of Pm/sup 1//sup 4//sup 8/ (T/sub 1//sub ///sub 2/ = 40.6 d) is of considerable reactor interest as a material produced from fission products at high fuel burnup. Neutron inelastic scattering measurements from methane were completed and compared with theoretical predictions based upon gas models. These measurements served to demonstrate the reliability of the experimental procedures and provided insight and valuable checks upon the theory as applied to a simple case. A previously unreported 3.14 hr isometric level was found in Y/sup 9//sup 0/ at 0.685 Mev. A program undertsken to determine the spin of the 512 kev level of decay of 2.3 d Np/sup 2// sup 3//sup 9/ by use of directional correlation measurements resulted in a preliminary level scheme. Continuing studies on gamma rays from the decay of both Pu/sup 2//sup 3//sup 9/ and Pu/sup 2//sup 4//sup 0/ resulted in gamma ray energy measurements not previously reported. A new end-window proportional counter was developed with improved geometry that minimizes the positive slope of the voltage plateau. An all-electronic scanning switch was developed to replace an existing 10 pole 100 position motor-driven rotary switch used to translate scaler data into a form usable by computer equipment. A number of new programs and computing techniques for the IBM-650 (and one for the IBM-704) were developed. These include a least squares program for use in calibrating the ARMF reguiating rod, a gamma spectrum interpolation scheme for use in generating analyzer response curves, an IBM-704 code for hot channel analysis in SPERT III, a nonlinear least squares program for the IBM-650, a program for use on the IBM-650 to edit input dath for the IBM-704 code PDQ, and a comprehensive MTR-ETR pricing and recordkeeping program for the irradiations in these reactors. (auth)
A survey of the Materials Testing Reactor shield by E Fast( Book )

2 editions published in 1953 in English and held by 9 WorldCat member libraries worldwide

Gamma intensities in the MTR gamma irradiation facility by T. R Wilson( Book )

4 editions published between 1954 and 1967 in English and held by 9 WorldCat member libraries worldwide

Finite representations of Bessel functions by D. D Dix( Book )

2 editions published in 1955 in English and held by 8 WorldCat member libraries worldwide

Pre-operational acceptance test procedures for the materials testing reactor by Phillips Petroleum Company( Book )

4 editions published between 1952 and 1956 in English and held by 8 WorldCat member libraries worldwide

Following the construction phase of the Materials Testing Reactor and the completion of Part I of the pre-operational acceptance tests, a series of acceptance tests on the reactor proper will be necessary. It is the purpose of this manual to describe in detail such a series of tests. These tests fall naturally into several well-defined phases. They are, approach to criticality; study of characteristics of the reactor and its components; establishment of power rating; and full power operation
Theory of power transients in the SPERT I reactor : final report by W. A Horning( Book )

4 editions published in 1958 in English and held by 8 WorldCat member libraries worldwide

Manual of radiochemical methods by D. G Olson( Book )

2 editions published in 1964 in English and held by 8 WorldCat member libraries worldwide

"The radiochemical methods used by the Analytical Branch, Atomic Energy Division of Phillips Petroleum Company for the determination of specific nuclides are described ..."
SPERT I destructive test program safety analysis report by A. H Spano( Book )

2 editions published in 1962 in English and held by 7 WorldCat member libraries worldwide

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 of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)
The dissolution of iron and nickel in dilute aqua regia by Richard Douglas Cannon( Book )

2 editions published in 1961 in English and held by 6 WorldCat member libraries worldwide

In laboratory studies the dissolution of iron in dilute nitric - hydrochloric acid mixtures shows an apparent reaction order of -0.62 with respect to HCl. No apparent order value for HNO/sub 3/ can be determined over the concentration ranges studied. Nickel dissolutions show apparent orders of 1.4 with respect to the HCl and 4.2 for HNO/sub 3/. Activation energies determined from 50 to 80 deg C are not constant, ranging from l0 to 20 kcal per mole for both metals. (arth)
Containment of iodine-131 released by the RaLa process by G. K Cederberg( Book )

2 editions published in 1961 in English and held by 6 WorldCat member libraries worldwide

Uncontrolled releases of large amounts of iodine occurred during the early stages of RaLa operation at the ldaho Chemical Processing Plant. A ten- fold reduction in the iodine content of the off-gas was achieved by process modifications, primarily the addition of mercury salts to the acidic process solutions. An additional ten-fold reduction was obtained by installing an activated charcoal adsorption unit in series with the original iodine removal scrubber. The iodine content of particulate entrainment limited the over-all iodine removal efficiency of the revised RaLa off-gas iodine removal system. (auth)
SPERT IV hazards summary report by F. L Bentzen( Book )

2 editions published in 1961 in English and held by 6 WorldCat member libraries worldwide

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)
A method of determining the intermediate energy neutron dose by Dale E Hankins( Book )

2 editions published in 1961 in English and held by 6 WorldCat member libraries worldwide

The intermediate energy neutron flux existing outside the biological shielding of reactors has not been studied to any great extent previous to this time, because of the lack of an instrument capable of detecting neutrons in the intermediate energy range. The instrument used at the MTR utilizes polyethylene spheres of various sizes to give different amounts of moderation and absorption to the impinging neutrons. A procedure for the approximate determination of the relative number of intermediate energy and fast neutrons is given. By graphical methods the following information is obtained: (1) fraction of intermediate neutrons, (2) fraction of fast neutrons, and (3) the approximate average energy of the fast neutrons. Since the instrument described can be used to determine the thermal neutron flux independent of intermediate and fast fluxes, only one instrument is required to determine the neutron flux in all three energy ranges. Dose calculations indicate the intermediate range neutrons give a dose greater than the dose delivered by fast neutrons around the MTR-ETR reactors under normal operating conditions. (auth)
Physical and operational features of a pulsed continuous countercurrent liquid-solids contactor by E. S Grimmett( Book )

2 editions published in 1961 in English and held by 6 WorldCat member libraries worldwide

The physical and operational features of a rapidly pulsed, continuous, countercurrent, liquid- solids contactor are described. The contactor consists of a 2-in.-dia. column containing five contact stages. Solids feeding, contactor operation, and pulsing methods are described. Ion exchange resin flow rates up to 150 lb/hr/ft/sup 2/ and calcined aluminum oxide flow rates up to 300 lb/hr/ft/ sup 2/ were obtained while feeding water at rates up to 2000 lb/hr/ft/sup 2/. Optimum stage design should extend these rates considerably. The effects of pulse frequency and amplitude upon the column flow rate are described. Column ion exchange efficiencies were determined for the systems Cu--Na and Cu--H/sub 2/ at a cationic concentration of 0.1N, Sulfate was the common ion in the Cu--Na system and nitrate ion in the Cu--H/sub 2/ system. Stage efficiencies of greater than 27 to 62% were obtained. HTU values are aiso reported. (auth)
Removal of tributyl phosphate and its degradation products from acidified uranyl nitrate solutions( Book )

2 editions published in 1964 in English and held by 6 WorldCat member libraries worldwide

A bench-scale natural-recirculation dissolver by E. E Erickson( Book )

2 editions published in 1962 in English and held by 5 WorldCat member libraries worldwide

A natural-recirculation dissolver closely approaching a stirred-tank reactor in behavior was developed, and its feasibility was demonstrated in the mercury-catalyzed dissolution of aluminum in nitric acid, It was designed to utilize the heat of reaction and evolution of gaseous reaction products for recirculation and mixing. The dissolvent was 5-molar nitric acid, containing mercury in concentrations from 10/sup -4/ to 10/sup -3/ molar as a catalyst. Dissolution rates for 1/7 scale, simulated MTR fuel elements of 2-S aluminum ranged from 0.25 to 4.0 mg/(cm/sup 2/)(rain), depending on the catalyst concentration in the dissolvent and the dissolvent flow rate. For comparable catalyst and acid feed concentrations, dissolution rates, based on unit net superficial velocities, were nearly 25 times those for dissolution of randomly- packed flat plates in a pilot plant, continuous, flooded dissolver. (auth)
SPERT program review by Warren E Nyer( Book )

2 editions published in 1960 in English and held by 5 WorldCat member libraries worldwide

The Reactor Safety Program is centered on those problems associated with the operation of nuclear reactors which involve the possibility of extreme hazards. The program involves studies of reactor dynamics, chemical reactions, and reactor containment. The earliest large-scale experimental work was the Borax series of tests. The principal objectives of the Borax series were related to the feasibility of boiling reactors but included the study of the self- limiting properties of such systems when subjected to sudden large additions of reactivity. Studies on the maximum step and ramp additions of reactivity that can safely be introduced into a few selected types of reactor cores were completed. Spert-II will be used to examine the influence of prompt neutron lifetime and to study the influence of special factors connected with the use of heawy water. Spert-III wiII be used to assess the importance of pressure and temperature, as well as other special factors, in power plants operating with boilingand pressurized-water reactors. Spert-IV will be used for investigating self-induced oscillations. Extensive changes are proposed in the Spert-I program. (W.L.H.)
Mixing and evaporation in a packed vessel by G. K Cederberg( Book )

2 editions published in 1961 in English and held by 5 WorldCat member libraries worldwide

In connection with an evaluation of the operability of a 36-inch diameter remote evaporator at the Idaho Chemical Processing Plant that was to be packed with a corrosionresistant neutron-poison packing for criticality control, an investigation in a 30-inch diameter vessel proved that air sparging effectively mixes solutions. The data showed that at similar spar;e rates the presence of the packing caused an increase in the time needed for complete mixing. The investigation showed that solutions are readily evaporated in spite of the presence of packing in the tank. (auth)
Preliminary evaluation of the 20% enriched uranium core for the Materials Testing Reactor by D. R DeBoisblanc( Book )

2 editions published in 1958 in English and held by 5 WorldCat member libraries worldwide

In November, 1957 the MTR was successfully operated on 20% enriched fuel for 646 Mwd, of which 550 were at the original MTR design power of 30 Mw. This power level was selected before the experiment on the basis of heat transfer limitations imposed by the use of a uniform fuel distribution. The plate spacing of the fuel elements deviated from specifications and it was necessary to select fuel elements and their placement in the core. One fuel elemert developed a pinhole in the cladding and released gaseous fission products soon level of the fission products did not necessitate shutting down the reactor, the defective fuel element was replaced as a precaution. No further difficulties occurred which could be attributed to the core. A number ef reactor physics measurements preceded the operation at power. The results of some of these measurements and other important items are listed. (auth)
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English (69)