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Rotating Machinery for Gas-Cooled Reactor Application

Description: From foreword: Representatives from various organizations met to discuss progress in the development of rotating machinery for gas-cooled reactors. The equipment covered included main blowers, shaft seals, gas turbines, gas bearing compressors, and other types of special compressors for reactor or experimental applications.
Date: June 1962
Partner: UNT Libraries Government Documents Department

NUMERICAL SOLUTION OF REACTOR STRESS PROBLEMS

Description: Generalized computer codes were devised for solving stress problems of some complexity. These codes were applied to stress problsms relating to the graphite moderator elements in the Experimental Gas-cooled Reactor. The stress relief obtained by aubdividing the moderator elements was evaluated. The distontion and bending moments of the elements were also determined. (auth)
Date: February 28, 1961
Creator: Redmond, R.F.; Hulbert, L.E. & Clark, R.W.
Partner: UNT Libraries Government Documents Department

EGCR CORE STRUCTURAL ANALYSIS. THE EFFECTS OF FAST-NEUTRON IRRADIATION AND THE BOWING CHARACTERISTICS OF THE GRAPHITE COLUMNS

Description: An analysis of the EGCR core structure was made to determine the lateral deflections (bowing) of the graphite columns resulting from shrinkage caused by fast-neutron irradiation, the life expectancy of each column due to restraints imposed on the bowing, and the reaction forces induced in the supporting structures. Based on currently avallable data for EGCR type graphite shrinkage and assuming experimental loop operation, a maximum bowing potential of 3.61 in. was calculated for an interior column. It was found that strains equivalent to the rupture strains observed from tensile tests could be expected after 4 to 6 years of full-power operation. Over half of the columns will reach these strains before the 20-yr reactor design life is reached. (auth)
Date: April 14, 1961
Creator: Moore, S.E. & Shaw, W.A.
Partner: UNT Libraries Government Documents Department

AXIAL TEMPERATURE AND PRESSURE DISTRIBUTION IN EGCR. DEVELOPMENT OF THE PTD-1 PROGRAM

Description: The equations and procedures of the PTD-1 program for obtaining the surface temperature distribution, the gas bulk temperature, and the pressure distribution for any of the 234 channels in the EGCR are developed. The program, in FORTRAN for the IBM 7090, is designed for any arbitrary power distribution yielding a constant gas outiet temperature or a given maximum surface temperature and includes the effects of thermal radiation on surface and bulk gas temperatures. The flow sheet and some results are included. (auth)
Date: February 18, 1964
Creator: Robinson, J.C. & Lence, J.T.
Partner: UNT Libraries Government Documents Department

NUMERICAL RESULTS FOR EGCR MODERATOR-ELEMENT STRESS PROBLEMS

Description: A detailed presentation is made of the thermal stresses calculated for the moderater elements in the Experimental Gas-Cooled Reactor. These results are discussed and some conclusions are presented. This report complements a previous report, BMI-1503, which defines the problems and discusses the methods of solution. (auth)
Date: July 1, 1961
Creator: Hulbert, L.E. & Redmond, R.F.
Partner: UNT Libraries Government Documents Department

Report of the Objectives and Plans for the AEC's Civilian Power Gas Cooled Reactor Program

Description: Progress in the U. S. civilian power gas-cooled reactor program is discussed. Gas reactors having technical features of high conversion ratio, high temperature, high fuel burnup, and capability of construction in large sizes make them very attractive as potential producers of economic power in the very near term. The operation of Peach Bottom-HTGR and EGCR in late 1964 and 1965, respectively, will contribute to the successful exploitation of thermal gas- cooled reactors. Since the graphite fuel concept promises very low fuel cycle costs along with reactor coolant conditions that can exceed current practice, it was concluded that the concept provides a long term potential that promises some very exciting possibilities. (auth)
Date: June 1, 1963
Creator: Pahler, R. E.
Partner: UNT Libraries Government Documents Department

DESIGN AND ANALYSIS: CORE STRUCTURE AND BOTTOM PLATE

Description: plate and to calculate its critical dimensions. Calculations were made which show that the plate and positioning keys meet the design requirements. Work sheets pertaining to the evaluation are included along with references to drawings. (J.R.D.)
Date: April 21, 1960
Creator: Goulden, P.V.
Partner: UNT Libraries Government Documents Department

LOSS-OF-PRESSURE ACCIDENT--HOT LEG PIPING FAILURE--QUICK-OPENING VALVES IN REACTOR BYPASS LINES

Description: Activities in a study to determine the differential pressure across the EGCR core following a pipe rupture in of quick-opening valves in bypass lines around the reactor are evaluated, and recommendations concerning these valves and other components are included. (J.R.D.)
Date: June 17, 1960
Creator: Landoni, J. A.
Partner: UNT Libraries Government Documents Department

EGCR Graphite Permeability Tests: Results of Forced Flow Experiments on Egcr Moderator-Grade Graphite

Description: Helium-permeability and porosity were determired at room temperature for specimens from a typical EGCR moderator-grade graphite block. Permeability, at a mean pressure of 2 atm, ranged from 26 to 200 (av. 86.5) millidarcys. Permeability data indicated that turbalent flow was never obtained with helium in these tests and that helium permeating the moderator graphite at EGCR operating conditions (taken to be: 600 deg C; DELTA P, 10 lb/in./sup 2/ per inch of graphite; mean P, 400 lb/in./sup 2/) was in the viscous flow region. Daroy's law and the reported constants are applicable for flow computations involving moderator graphite under these conditions. Porosity ranged from 20.6 to 29.4% (av. 23.8%), and there was no correlation between porosity and pemaesbility variations. The large variations encountered were believed to reflect the nonuniformity of the specimens, since duplicate determinations showed excellent agreement. Permeabilfty did not change appreciably with direction of flow and did not vary consistently with respect to the extrusion or any other axis. Preparation of the specimens did not appear to introduce appreciable surface effects. (auth)
Date: March 24, 1961
Creator: Ward, W. T. & Truitt, J.
Partner: UNT Libraries Government Documents Department

LOSS-OF-PRESSURE ACCIDENT. HOT LEG PIPING FAILURE

Description: Results are presented of a study to determine if a circumferential rupture in the EGCR hot-leg piping just downstream of the block valve, accompanied by complete lateral displacement of the pipe ends such that the exhaust flows do not oppose each other, will result in excessive pressure differentials across the reactor core. Conclusions and recommendations are included. (J.R.D.)
Date: January 11, 1962
Creator: De Agazio, A.W.
Partner: UNT Libraries Government Documents Department

Numerical Three Dimensional Temperature Analysis of the EGCR Fuel Rod

Description: The temperature field in an outer rod of the Experimental Gas Cooled Reactor fuel assembly is presented. The numerical analysis has taken account of the three dimenhas considered the four resons involved. that is. fuel pellets, ceramic disc, steel end caps. and cladding. The convective heat transfer coefficient and fluid temperature were considered constant over the 2.185 in. of rod length studied. (auth)
Date: March 21, 1960
Creator: Epel, L. G.
Partner: UNT Libraries Government Documents Department

COOLANT FLOW DISTRIBUTION PERMEATION OF COOLANT THROUGH GRAPHITE

Description: An investigation was conducted to determine the effects of He leakage through EGCR moderator columns and fuel element sleeves on the core coolant flow distribution reported in study II-179-1. Results of the investigation are discussed and recommendations for design changes so that the total coolant bypass will remain the same as previously reported are made. (J.R.D.)
Date: April 1, 1960
Creator: Kintner, L.L.
Partner: UNT Libraries Government Documents Department

TEMPERATURE DISTRIBUTION, MODERATOR AND REFLECTOR REACTOR CORE

Description: Studies were made to determine revised moderator coolant flow requirements in the EGCR core for the latest design. The temperature distribution in the graphite columns was also determined. The total moderator coolant flow was calculated to be 24,024 lb/hr and 31,020 lb/hr for full power and maximum anticipated power, respectively. It was concluded that the maximum moderator temperature (1220 deg F average over the cross section), at full power operation, occurs near the peak heat flux at a location 6.25 to 7.25 ft from the bottom of the active core. The temperature in most of the graphite columns varies from 555 to 1150 deg F over the lower half of the column and 1150 to 990 deg F over the top half of the column. The maximum surface temperature is less than 1110 deg F. Temperature distributions are shown on graphs. (M.C.G.)
Date: December 21, 1960
Creator: Cheng, F.
Partner: UNT Libraries Government Documents Department

GAS-COOLED REACTOR PROGRAM. QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JUNE 30, 1961

Description: Activities are discussed for research in design investigations, and materials development and testing conducted in connection with the development of the EGCR. The discussions are given in terms of: reactor physics; reactor design studies; heat transfer and fluid now investigations; materials development; in- pile and out-of-pile testing of components and materials; and development of test loops and components. (B.O.G.)
Date: August 28, 1961
Partner: UNT Libraries Government Documents Department

GAS-COOLED REACTOR PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING SEPTEMBER 30, 1959

Description: ; D = > 6 ; : 6 < 7 8 9 7 9sion theory calculation of the power-density distribution in the EGCR was made in order to reduce the uncertainties in the results of previous calculations. A comparison was made of calculated neutron flux distributions in seven-rod fuel clusters and the preliminary results obtained in the Physical Constants Test Reactor at Harford. Neutron flux ratios for the EGCR lattice cell were calculated for fuel enrichments of 2.0 and 2.6%. Studies were made of the power densities attainable in gas-cooled reactors operated with ceramic material as both fuel and moderator. Extensive studies were conducted to determine how the multiplication factor of a gas-cooled reactor varies with the number of rods in the fuel element cluster, cladding thickness, cladding material, inter-rod spacing, lattice pitch, solid and cored rods, fuel enrichment, and fuel-tomoderator ratio. Reactor Design Studiesi A theoretical study is being conducted of the deflections and stresses in fuel element cladding based on arbitrary temperature distributions. Tests were run to determine the load-carrying ability of the threaded-pin-type graphite column joints proposed for the EGCR. A test program was initiated for studying the behavior of metal-clad graphite bodies. An analytical model for investigating temperature structure and thermal stability of a seven rod fuel element cluster was developed. Calculations were made of the nataral frequencies aud an amplitude of vibrations in the EGCR fuel element clusters. A parametric study was completed from which the diameter of the coolant channel, pressure drop across the core, and pumping power can be evaluated for a helium-cooled reactor fueled with UO/sub 2/ clad with stainless steel. A detailed analysis was made of the He purification system needed for limiting graphite bunnup and carbon mass transfer in the EGCR. In a design study of advanced steam generators, data ...
Date: December 1, 1959
Partner: UNT Libraries Government Documents Department