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CRITICAL CONCENTRATIONS FOR HRT-TYPE REACTORS SUBJECTED TO VARIOUS CONDITIONS

Description: BS>Critical concentration calculations were made for several D/sub 2/O-H/ sub 2/O moderated HRT-type reactors with 30- and 28-in. core diameters and pressure vessel diameters of 60 and 54 in. A core temperature of 300 C was assumed for all cases while the blanket temperatures assumed the values 250, 280, and 300 C. The assumed moderator compositions were 80, 90, and 100% D/sub 2/O. (auth)
Date: February 1, 1959
Creator: Chalkley, R.
Partner: UNT Libraries Government Documents Department

Spectral Shift Control Reactor Basic Physics Program. Measurement and Analysis of Perturbed Lattices of Slightly Enriched Uo$Sub 2$ Moderated by D$Sub 2$O-H$Sub 2$O Mixtures

Description: Critical experiments and theoretical calculations of reactivity and flux perturbations were performed. Three major critical assemblies were studied; the basic cores had 1.206-cm diameter, aluminum clad, 2.46%-enriched UO/sub 2/ fuel rods on a square lattice pitch of 1.511 cm. Moderator compositions for the three major cores had 0, 50, and 72 mole% D/sub 2/O. Bonic acid was added to keep the core radius constant at 61.11 cm. Data on a zone-loaded core with an inner zone of 93%-enriched aluminum clad UO/sub 2/-ThO/sub 2/ fuel rods surrounded by an outer zone of 4%-enriched stainless steel clad UO/sub 2/ fuel rods are also included. The moderator composition was 81% D/sub 2/O. Perturbers were introduced axially by removing fuel pins to create moderator gaps and by inserting perturbing blades and cruciform rods. The blades were 10 inches wide. The cruciplease delete the above abstract=====
Date: October 1, 1963
Creator: Engelder, T. C.; Clark, R. H.; DeRoche, E. J.; Fairburn, G. T.; Hallam, J. W.; Hostetler, D. R. et al.
Partner: UNT Libraries Government Documents Department

PRELIMINARY REPORT ON HEAT GENERATION AND STRESSES IN THE WALL OF A SPHERICAL HRE-4 PRESSURE VESSEL

Description: The heat geueration and thermal stresses in spherical HRE-4 vessels 3 to 4 1/2 ft in diameter with clad and solid stainless steel walls were invegtigated. Parameters included thorium slurry concentrations and moderator material (D/sub 2/ O and H/sub 2/O). The prirnary purpose of this study was to determine the influence of thermal stresses on the selection of the core size for the HRE4 reactor. Curves are presented which facilitate relatively rapid determination of stresses for the range of vessels considered. It is concluded that steady- state thermal stresses in the clad or solid stainless steel vessels considered will not have to be a determining factor in the selection of a core size, provided the power density does not exceed approximately 15 kw/l in the clad vessels and 8 kw/l in the solid stainless steel vessels. (auth)
Date: December 31, 1958
Creator: Cheverton, R.D.
Partner: UNT Libraries Government Documents Department

A PRELIMINARY STUDY OF THE NUCLEAR STABILITY OF FLUIDIZED BED REACTORS

Description: A preliminary study was made of the nuclear stability of reactors in which particles containing U/sup 233/ and Th/sup 232/ are fluidized by liquid water. D/sub 2/O moderated reactors were found to be least sensitive to changes in bed height at thorium concentrations in the range 80 to 120 g/liter, H/sub 2/O reactors in the rarnge 1500 io 2400 g Th/ liter, and mixed H/sub 2/O-D/sub 2/O reactors at thorium concentrations between those for pure moderators. A reactor operated at the thorium concentration at which it is least sensitive to change in bed height was found to be stable in response to a cyclical variation in height. (auth)
Date: January 12, 1959
Creator: Benumof, R. & Rosenthal, M.W.
Partner: UNT Libraries Government Documents Department

BRIC-AN IBM-704 TWO-DIMENSIONAL NUCLEAR-THERMAL DEPLETION PROGRAM WITH DISTRIBUTED VOID EFFECTS

Description: BRIC is a two-dimensional nuclear-thermal depletion program to study the effects of moderator boiling on the neutron flux distribution and the depletion of the timedependent isotopes. The program requires an IBM-704 with a memory of 32,768 words and ten tape units. (auth)
Date: January 1, 1961
Creator: Jacobi, W.M.; Lawton, T.J.; Meanor, S.H. & Parrette, J.R.
Partner: UNT Libraries Government Documents Department

Physics Calculations for Control Rods in the First Yankee Core

Description: BS>The calculation of control rod worths in light water moderated reactors is presented. Considerable experimental information from the Yankee critical experiments is discussed and applied to improve the theory. The calculation system thus deduced from theory and experiment is then applied to the final design of the first Yankee core to provide a prediction of the shutdown available in its control rods. These calculations show that the 24 control rods presently provided are sufficient to shut the hot, clean, zero power first core down by 3%. (auth)
Date: September 1, 1959
Creator: Arnold, W. H., Jr.
Partner: UNT Libraries Government Documents Department

Nuclear Instrumentation System System Description No. 15

Description: Modifications are described in the nuclear instrumentation system of the Shippingport Pressurized Water Reactor designed to monitor neutron flux level from source to 150% of reactor designed full power output, to compute the rate of change of the neutron flux level for indication and control, and to provide level and rate of change of level signals for reactor shutdown to prevent reactor damage. (C.H.)
Date: May 12, 1956
Creator: Wilson, N. E.
Partner: UNT Libraries Government Documents Department

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.
Partner: UNT Libraries Government Documents Department

MODIFICATION OF THE EXPERIMENTAL BOILING WATER REACTOR (EBWR) FOR HIGHER- POWER OPERATION

Description: 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.
Partner: UNT Libraries Government Documents Department

A LOW COST PHYSICS AND ENGINEERING TRAINING REACTOR. Reactor Design and Feasibility Study

Description: The conceptual design of a low cost training reactor for the instruction of physicists and engineers is covered. It is conceived as an instructional tool for a course such as that given at the Oak Ridge School of Reactor Technology. The reactor is of a modified pool type, and is designed for a maximum power level of one Mw. This arrangement will accommodate engineering experiments, shielding experiments, and critical experiments as well as being useful as a neutron and gamma source. (auth)
Date: August 1, 1956
Creator: Brown, R.A.; Ackerman, M.W.; Batter, J.F. Jr.; Bell, D.W.; Dunlay, J.B.; Ellis, G.K. et al.
Partner: UNT Libraries Government Documents Department

FEASIBILITY AND CONCEPTUAL DESIGN FOR THE STEP LOSS OF COOLANT FACILITY

Description: A summary of studies conducted on the pressurizedwater reactor loss-of- coolant accident is presented, and an experimental safety program is proposed. The various phenomena involved in the loss-of-coolant accident, related research and development programs, assumptions currently used to predict the various physical phenomena, and the general approach to be used in conducting the safety tests are discussed. In order to accomplish the loss-of-coolant experimental safety program, a dry containment test facility is proposed for construction at the Test Area North of the National Reactor Testing Station in Idaho. The site selection utilizes existing support facilities suited for performing nuclear safety tests requiring experiment assembly areas and post-test analytical examination of the irradiated nuclear components. (auth)
Date: April 24, 1963
Creator: Wilson, T.R.; Hauge, O.M. & Matheney, G.B.
Partner: UNT Libraries Government Documents Department