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Modified Purification System Performance Test. Core 1, Seed 2. Test Results T-641124-A. Section 2

Description: The effectiveness of purification in controlling plant radioactivation rates was determined through comparison of the actual effects produced from plant operation with and without purification. With the purification system in service, there was no appreciable difficulty in maintaining the reactor coolant within reference water specifications. In addition, there was no discernible increase of crud deposition in the coolant system as determined by direct radiation measurements of the purification… more
Date: July 21, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

ORR Operations for Period April 1961 to April 1962

Description: A summary of the activities in the 4th operational year is presented for the ORR. On-stream time at 30 Mw was relatively steady with 75.4 and 83.7% representing the lowest and highest quarters. Modification of the ball-latch mechanism of the shim-rod-drives is in progress. The primary cooling system bypass control valve was modified and a d-c pony motor was added at the No. 3 primary pump. This addition was made to increase the reliability of adequate water flow for afterheat cooling. A study o… more
Date: October 16, 1962
Creator: Binford, F.T.; Casto, W.R. & Colomb, A.L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Measurements and Changes on SM-1 Core II During Period October 1, 1961 to May 30, 1962

Description: Tests at the SM-1 reactor are reported for the period October 1, 1961, to May 31, 1962. Loading changes were made in SM-1 Core II during the scheduled semiannual shutdowns in October to November 1961 and April to May 1962. Core physics tests include control rod bank calibrations, bank position at several temperature and xenon poison conditions vs core changes and energy release, shutdown neutron source decay and startup channel testing, and critical rod positions for stuck rod configurations. S… more
Date: July 1, 1962
Creator: Motte, F. G.; Best, W. C. & Kortheuer, J. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Detailed Stress Analysis of SM-1 Steam Generator Tube Sheet

Description: The detailed stress analysis of the SM-1 steam generator tube sheet showed it to be safe from strain cycling damage. However, the pressure stresses were greater than the yield strength during the hydrostatic test. The differential between pressure stresses and yield strength indicates that some initial deformation may have taken place in the tube sheet. (auth)
Date: July 11, 1962
Creator: Busuttil, J. J. & Chittum, R. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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. Varia… more
Date: July 1, 1961
Creator: Christensen, H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

VOID COEFFICIENT OF REACTIVITY ASSOCIATED WITH THE ISLAND REGION OF THE HFIR

Description: Changes in neutron multiplication caused by voids in the island of the HFlR were calculated and measured experimentally. The results indicated that with only water initially in the island the maximum change in neutron multiplication ( DELTA k/sub max) associated with island voids is 0.032 with a corresponding void fraction of 70%. With a simulated 300 g Pu target in the island DELTA k/sub max/ was 0.0l6, and the corresponding void fraction was 42%. In view of these large changes in neutron mult… more
Date: November 15, 1961
Creator: Cheverton, R.D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

An Evaluation of Reactor Concepts for Use as Separate Steam Superheaters

Description: Various reactor concepts were compared for use as a separate superheater which could be added on to an advanced 300-Mwe reactor producing saturated steam. Fossil steam plant superheat temperatures were used as a criterion for selecting nuclear superheat temperatures. Therefore, the performance specified for the superheater was a minimum exit steam temperature of 566 deg C (1050 deg F) when supplied with saturated steam at either 7l atm (l050 psia) or l67 atm (2450 psia). A preliminary screening… more
Date: October 1, 1961
Creator: Lennox, D. H.; MacFarlane, D. R.; Brubaker, R. C.; Martinec, E. L.; Rohde, R. R.; Toppel, B. J. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Reactor Controls Reliability and Maintenance at the ORR

Description: Early evaluation and study of the various criteria for the ORR clearly established the requirements for the fundarnental safety and control instrumentation, and the instrumentation was well integrated into the design and construction of the plant. However, it was not designed for the optimum of maintenance convenience and minimum reactor down time in the event of instrument failure. The revisions and additions that have been made in the physical plant and maintenance which have resulted in a de… more
Date: August 28, 1962
Creator: West, K.W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Nuclear Safety Program Semiannual Progress Report for Period Ending June 30, 1962

Description: Progress in nuclear safety research and development is reported. Topics covered include: reactivity effects of fuel displacements in a pool-type reactor, release of fission products on out-of-pile melting of reactor fuels, release of fission products on in-pile melting of reactor fuels, fission product transport evaluations, characterization and control of accident-released fission products, nuclear safety pilot plant, preparation of reactor containment hardbook, and radiochemical plant safety … more
Date: August 24, 1962
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Periodic Waste Disposal System Material Balance Test. Core I, Seed 1. Test Results (T-641317). Section 3

Description: An investigation was conducted to determine the adequacy of storage capacity and operating procedures of the waste disposal system during reactor cooldown for refueling. It was found that the waste storage capacity and the operating procedures were adequate. (J.R.D.)
Date: October 17, 1960
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

CHARGING AND DISCHARGING OF DEMINERALIZER RESINS. CORE I, SEED 2. Test Results (T-612085). Section 1

Description: An investigation was conducted to flush the 1BD coolant purification system demineralizer of resin and to measurc the radiation level at pornts on the resin discharge line. It was found that the system demineralizer was satisfactorily flushed of resin. It was also found that the amount removed is measured by the amount required to recharge the demineralizer. (J.R.D.)
Date: January 10, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Mathematical Analysis of Rippling of Type 1 Fuel Plates. Part 1

Description: Rippling phenomena due to heating in fuel plates of SM and PM type reactors are investigated analytically using small deflection theory of plates. Temperature variations across the width of the plate are accounted for. Detailed calculations are conducted for simply supported plates. It is found that within the limitations imposed by small deflection theory that the amplitude of the plate ripples induced by the heating is directly proportional to the initial amplitude. (auth)
Date: July 1, 1962
Creator: Beck, S. D. & Miller, J. V.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Pathfinder Atomic Power Plant Analog Simulator.

Description: The steady-state and dynamlc characterlstics of the Pathfinder systems were simulated on a dual-console Pace Analog Computer. The derivation of the system equations, the assumptions involved, their numerical evaluation, and their conversion to analog circuitry are described. The Pathfinder system is unique because of the internal nuclear superheater, and controlled, forced recirculation. In simulating this system the following major assumptions were maden the reactor is a boiling-water type; mo… more
Date: January 31, 1962
Creator: Crimmins, D. H.; Mohr, D. & Stone, J. T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Primary Shield Optimization Survey for the PL-3 Reactor

Description: A detailed study of four reactor and shield configurations was made. Two basic reactor types, the boiling water and pressurized water reactors were considered. Shield materials of lead-water and iron-water were used with varying thicknesses for determining the optimum shield configuration for the PL-3 reactor. Also presented is a survey of available shielding codes. (auth)
Date: June 28, 1962
Creator: Scoles, J. F. & Crouch, A. N.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Experimental and Analytical Reactivity Studies of Clean Critical Stainless Steel Cores

Description: ABS>The results are presented of critical water height measurements made on close-packed lattices of Spert III, highly enriched, plate-type, stainless- steel-clad fuel elements. Experiments were conducted with cores containing no control rods and with cores containing a single, fully-inserted control rod. The "clean critical" data obtained in these experiments were used to test the validity of various aspects of a four-group, diffusion theory analysis of the full scale Spert III reactor. The re… more
Date: June 16, 1961
Creator: Spano, A. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Dynamic Simulation of Multi-Pass Pressurized Water Nuclear Power Plants by Analog Computer Techniques

Description: A kinetic model of the primary loop of a multi-pass pressurized water reactor power plant is developed to evaluate, by analog computer techniques, the transient response characteristics under conditions of steam generator load and reactor control rod perturbations. Using the 2-pass 28 Mw(t) SM-2 reactor as a typical plant, transient behavior patterns are illustrated and examined for a variety of load inputs, variations in plant constants, and analog model simplifications. (auth)
Date: June 1, 1961
Creator: Brondel, J. O.
Item Type: Refine your search to only Thesis or Dissertation
Partner: UNT Libraries Government Documents Department
open access

IN-PILE GAS-COOLED FUEL ELEMENT TEST FACILITY

Description: Paper presented at American Nuclear Society Meeting, June I8-21, 1962, Boston, Mass. Design and operating problems of unclad and ceramic gas-cooled reactor fuels in high temperature circulating gas systems will be studied using a test facility now nearing completion at the Oak Ridge Research Reactor. A shielded air-tight cell houses a closed circuit gas system equipped for dealing with fission products circulating in the gas. Experiments can be conducted on fuel element performance and stabilit… more
Date: July 10, 1962
Creator: Zasler, J.; Huntley, W. R.; Gnadt, P. A. & Kress, T. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

THE OAK RIDGE RESEARCH REACTOR (ORR), THE LOW-INTENSITY TESTING REACTOR (LITR), AND THE OAK RIDGE GRAPHITE REACTOR (OGR) AS EXPERIMENT FACILITIES

Description: >Characteristics of the ORR, LITR, and OGR that experimenters have found to be important are listed. The results of a survey conducted among experimenters on the utility of the reactors for various types of experiments are discussed, and some changes which might be made to improve the utilization are listed. A brief outline, with references, of most of the experiments currently being performed is included. (auth)
Date: August 28, 1962
Creator: George, K.D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Analysis of the Initial Nuclear Superheat Critical Experiments. Supplementary Study Related to Bonus and Nuclear Superheat Programs

Description: A critical experiment program is carried out in a configuration similar to the BONUS reactor. The results give information concerning: the effects of different boilersuperheater geometries; the reactivity changes associated with superheater voiding or flooding; power regulation between the boiler and superheater regions; epithermal transmission probabilities for B-stainless steel and Cd control rods; the power flattening characteristics; and void simulation properties. The calculational methods… more
Date: January 30, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Modified Purification System Performance Test. Core I, Seed 2. Test Results (T-641124). Section 1

Description: An investigation was conducted to establish an operating procedure for flushing water through a purification system demineralizer which was out of service for an extended period, and to determine the demineralizer serviceability. (J.R.D.)
Date: September 14, 1960
Creator: Duquesne Light Company
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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