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Description: The hazards evaluation was modified to reflect certain changes made to the equipment as a result of operating experience. These changes included: the addition of a startup interlock circuit; the modification of a startup interlock circuit; several minor modifications to the control rod actuators; and the addition of the tube-sheet cooling system. (M.C.G.)
Date: May 1, 1960
Partner: UNT Libraries Government Documents Department


Description: The GCRE-I hazard summary report is supplemented in the following areas: geometry and operation of the steam cooling system, the reactor coolant by-pass, and by-pass valving; the means by which by-passed circuits are prevented from remaining unintentionally disabled; design details, and details of procedure for core flooding operations. (A.C.)
Date: March 1, 1959
Partner: UNT Libraries Government Documents Department


Description: The evaluation of the data generated during the full power and limited endurance tests of the ML-1 mobile nuclear power plant indicates that the reactor performs in accordance with the design specifications. During the 101 hr test period, the reactor attained a maximum power of 3.44 Mw( and 247 kw(e) was measured at the output shaft of the turbine-compressor set. No operating limits were exceeded during these tests and all systems performed satisfactorily Except for the known performance deficiency of the turbinecompressor set, which prevented the attainment of design output power, no operational, stability, or control problems were encountered. All test objectives were achieved and the tests were considered completely successful. (auth)
Date: July 1, 1963
Creator: Kattchee, N.
Partner: UNT Libraries Government Documents Department

ARMY GAS-COOLED REACTOR SYSTEMS PROGRAM. Monthly Progress Report for April 1959

Description: The Army Gas Cooled Reactor System Program includes water moderated heterogeneous reactor (Gas Cooled Reactor Experiment I), a graphite moderated homogeneous reactor (Gas Cooled Reactor Experiment II), a mobile gas cooled reactor (ML-1), and the co ordination of thc Gas Turbine Test Facility. The progress of each project, the associated tests and data evaluation, the applicabie design criteria, and the fabrication of reactor components are briefly summarized. (For preceding period see IDO-28538.) (W.D.M.)
Date: May 25, 1959
Partner: UNT Libraries Government Documents Department

Army Gas-Cooled Reactor Systems program: alternator final design report

Description: The development and testing of a demonstration brushless alternator for the ML-1 mobile nuclear power plant is described. The brushless concept was selected after it became apparent that a conventional power generator could not satisfy the ML-1 weight and size requirements. The demonstration alternator fabricated and tested under this program did not meet all performance specifications; the efficiency was low and the unit could not be operated for significant periods of time without overheating. However, a large body of useful data was accumulated during the extensive development program. Of special interest are data on the rotor and stator design, the cooling requirements and on the distribution of eddy current losses. Analysis of the data indicates that a brushless alternator, only slightly larger and heavier than was specified for the ML-1, could be developed with a modest additional effort.
Date: June 1, 1964
Partner: UNT Libraries Government Documents Department

Effect of high temperature in-pile irradiation upon high density UO/sub 2/ bodies

Description: As a part of the ML-1 fuel evaluation program, high density bulk UO/sub 2/ pellets were irradiated to burnups equivalent to 8500 hours of ML-1 reactor operation (3.0 a/o U-235) under temperature conditions approximating those expected in the reactor (1750/sup 0/F peak cladding surface temperature). Post irradiation examination revealed severe fuel cracking although fission product gas release measurements agreed well with calculated values. No structural changes were revealed by metallography and X-ray diffraction except for a small unidentified phase observed at the fuel grain boundaries. This phase is considered to be the result of impurities or fission product gas sites. The data obtained on UO/sub 2/ fuel is in agreement with that determined in irradiation experiments performed at other facilities. The performance of UO/sub 2/ fuel was sufficiently good to permit use in the ML-1 reactor.
Date: January 1, 1977
Creator: Saling, J. W. & Titus, G. W.
Partner: UNT Libraries Government Documents Department


Description: Investigatiors were made of various materials for development of metal- canned and semi-homogeneous GCRE-II fuel element concepts. The materials were studied for application to development of fuels, grapanite, silicon-silicon carbide coatings, metal claddings, carburization barrier coatings, and graphite joining. A survey of the literature showad that uranium carbide fuels are superior to other types for the applications described and that refractory metal or metal carbide fuel coatings appear superior to other types for use with the types of graphite investigated. Experimental measurements were made of the thermal conductivity, tensile strength, stress-strain reiationships, and thermal expansion of graphite powdsr bonded with baked carbon at a final firing temperature of 760 deg C. Results showed that these materials were stronger and more isotropic at all test temperatures than a standard structure graphite such as ATJ. The thermal conductivity is somewhat lower and the thermal extansion slightly higher than the corresponding properties of ATJ. A silicon-silicon carbide coating was developed as an osidation-resistant coating for graphite. Preliminary air oxidation tests at 1000 deg C showed that the first samples survived 2000 hr with 10% failure. Subsequent experiments showed that it is reasonable to expect better performance in further tests. Tests for compatibility with graphite were conducted on zirconium, Zircaloy-2, "A" nickel, and K-Monel at 1750 and 1850 deg F for 1000 and 1500 hr. Chemical analyses, metallography, and tensile tests indicated that the K-Monel is the material most compatible with graphite; it possesses good strength and ductility with negligible carburization or carbon diffusion. Zircaloy-2 tubing showed a growth of from 3.4 to 3.8% when thermal cycled 100 times between 850 and 1850 deg F. Tests for compatibility with Hastelloy X were conducted on graphite samples coated with molybdenum, niobium carbide, and zirconium carbide at 1750 deg F and 300 psi for 1000 and ...
Date: December 30, 1960
Creator: Carpenter, R. & Del Grosso, A.
Partner: UNT Libraries Government Documents Department

RUBIDIUM AND CESIUM EVALUATION PROGRAM. SPACE POWER SYSTEMS TECHNOLOGY STUDIES. Quarterly Technical Report for Period February 1 through April 30, 1961. Report No. 1

Description: A summary is given of the development of equipment to obtain data for the corrosion and solubility properties of rubidium, and for the corrosion and thermodynamic properties of cesium. (B.O.G.)
Date: October 31, 1961
Creator: Young, P.F.
Partner: UNT Libraries Government Documents Department

SPACE POWER SYSTEMS TECHNOLOGY STUDIES. RUBIDIUM EVALUATION PROGRAM. Report No. 12. Quarterly Technical Progress Report for Period November 1, 1960 through January 31, 1961

Description: A stainless steel loop test with liquid rubidium was run for 172 hr, and the design temperature of 1510 deg F at approximately 15% vapor quality rubidium in the boiling phase was achieved. The problems of loop operation are discussed, e.g., trapped gas bubbles and argon leakage from the argon-to-water heat exchanger. The rubidium was made to boil, and its boiling point was determined to be 1325 deg F at 33 psia and 1450 deg F at 60 psia. The discrepancy between measured and literature boiling points is probably due to the fact that the thermocouples did not measure the actual rubidium temperature. The density of rubidium was measured at several temperatures from 175 to 1340 deg F and compared with literature values. (D.L.C.)
Date: October 31, 1961
Partner: UNT Libraries Government Documents Department


Description: The corrosion and heat transfer characteristics of boiling sulfur were studied in order to evaluate the feasibility of using a boiling sulfur cycle to extract energy from a reactor. Work on the program included a literature survey of corrosion studies, capsule corrosion tests, dynamic loop corrosion tests, and boiling heat transfer experiments. (auth)
Date: December 1, 1960
Creator: Sawle, D.R.
Partner: UNT Libraries Government Documents Department

Army gas-cooled reactor systems program. ML-1 pressure vessel technology evaluation. Summary report

Description: A limited program of evaluation of the ML-1 technology as related to the design of the calandria/pressure vessel assembly was completed. This work included: a photoelastic experimental program which resulted in the conclusion that values recommended by Langer and O'Donnell for the effective elastic constants employed in the analysis of ligament stresses in the ML-1 type tube sheets in the past results in a good approximation of the bending constants but in a value of the tensile elastic modulus which is about 40% too high; several modifications to the TSA computer code for stress calculations which improved the precision and flexibility of the code; and a very preliminary evaluation of the properties of candidate materials for use in a high-performance, advanced ML-1 type pressure vessel.
Date: July 1, 1965
Creator: Eggert, W.K. & English, W.F.
Partner: UNT Libraries Government Documents Department


Description: As a part of the ML-1 reactor fuel evaluation program, UO/sub 2/-- BeO bodies containing two UO/sub 2/ compositions, 70 and 80 wt% UO/sub 2/, were irradiated to a burnup equivalent to 8750 hours of ML-1 reactor operation (8.5% U/ sup 235/). It was estimated that maximum cladding surface temperatures of 1710 nif- F were attained during irradiation. Reference design burnup for the ML-1 is 9.7% U/sup 235/, in 10,000 hours of operation at maximum clad surface temperatures of 1750 nif- F. Postirradiation examination of the test specimens revealed that the cladding of one of the 80 wt% UO/sub 2/-- BeO specimens had ruptured after severe swelling. All other specimens showed little external effect from the irradiation. Fission gas release from the fuel varied between 0.59 and 2.7% except for the failed specimen which released about 69%. Considerable change was observed in the microstructure of the irradiated specimens although subsequent x-ray diffraction examination did not indicate serious damage to the crystal structure of either the BeO or UO/sub 2/. The data obtained from this experiment are in substantial agreement with that determined in irradiation experiments performed at other facilities. The performance of 70 wt% UO/sub 2/--BeO was considered to be suitable for use in the ML-1 reactor. (auth)
Date: January 1, 1963
Creator: Titus, G.W. & Saling, J.H.
Partner: UNT Libraries Government Documents Department