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Further Critical Experiments on a Small Reactor of Enriched U-235 with Al-H₂O Moderator and Beryllium Reflector

Description: From introduction: "The purpose of the present report is to describe briefly certain improvements that have been made in the experimental facilities and to present the results of further measurements on critical masses and spatial neutron flax distributions in Be reflected reactors having square and thin slab geometries."
Date: August 17, 1948
Creator: Martin, A. B. & Mann, M. M.
Partner: UNT Libraries Government Documents Department

The Time-Dependent Neutron Flux for a Pulsed Subcritical Assembly

Description: Report discussing a study regarding the time-dependent neutron flux for a pulsed subcritical assembly. From abstract: "A two-group analysis has been applied to the transient behavior of the thermal flux in a subcritical assembly which has been given a short burst of fast neutrons."
Date: April 15, 1957
Creator: Fultz, Stanley C.
Partner: UNT Libraries Government Documents Department

Feasibility Study of a Neutron Flux Bolometer

Description: Abstract: "A small low impedance flux monitoring device has been built which utilizes the heating (and consequent chance in resistance) of a boron coated platinum wire. Feasibility tests in the Argonne-GP-3 reactor indicate that it may develop into a wide range instrument, particularly useful at high flux levels. Design and constructional details of the bolometer and its associated circuits are given. The circuitry is somewhat unique since it is in effect an all electronic self-balancing bridge. Future lines of research are suggested to solve some of the remaining problems."
Date: September 5, 1951
Creator: Leonard, Robert R.
Partner: UNT Libraries Government Documents Department

Neutron Flux and Spectra Measurements in the Sandia Pulsed Reactor Facility (SPRF)

Description: Introduction: Neutron measurements were made in the pulsed reactor building and on the safety screen of the pulsed reactor in order to determine the neutron yield of the reactor as a function of (1) distance from the reactor centerline, (2) direction in the reactor building, and (3) position on the reactor safety screen.
Date: January 1962
Creator: Buckalew, William H.
Partner: UNT Libraries Government Documents Department

A Problem in the Conversion of Neutron Dose to Flux

Description: This report discusses the calculation involved in converting neutron dose to flux. In order to test conversion calculation methods, the flux entering the Oak Ridge National Laboratory Lid Tank and the water by the north face of the Oak Ridge National Laboratory Bulk Shielding Reactor was used.
Date: June 29, 1953
Creator: McMillan, D. E.
Partner: UNT Libraries Government Documents Department

SRE Instrumentation and Control

Description: Introduction: This memo gives a general description of the components and equipment affecting the control of the SRE, and the equipment associated with all reactor services.
Date: May 21, 1956
Creator: Hall, R. J.
Partner: UNT Libraries Government Documents Department

Studies of Axial-Leakage Simulations for Homogeneous and Heterogeneous EBR-II Core Configurations

Description: When calculations of flux are done in less than three dimensions, leakage-absorption cross sections are normally used to model leakages (flows) in the dimensions for which the flux is not calculated. Since the neutron flux is axially dependent, the leakages, and hence the leakage-absorption cross sections, are also axially dependent. Therefore, to obtain axial flux profiles (or reaction rates) for individual subassemblies, an XY-geometry calculation delineating each subassembly has to be done at several axial heights with space- and energy-dependent leakage-absorption cross sections that are appropriate for each height. This report discusses homogeneous and heterogeneous XY-geometry calculations at various axial locations and using several differing assumptions for the calculation of the leakage-absorption cross section. The positive (outward) leakage-absorption cross sections are modeled as actual leakage absorptions, but the negative (inward) leakage-absorption cross sections are modeled as either negative leakage absorptions (+-B² method) or positive downscatter cross sections (the ..sigma../sub s/(1 ..-->.. g) method).
Date: August 1985
Creator: Grimm, K. N. & Meneghetti, D.
Partner: UNT Libraries Government Documents Department

Instrumentation and Some Related Problems for Neutron Flux Measurement of the Los Alamos Scientific Laboratory Kiwi-A Reactor

Description: Abstract: "A description of the instrumentation for neutron flux measurement from Kiwi-A at NTS is given as some discussion of start-up considerations, detector design, system response adjustments, and power calibration of the reactor at some low level of power."
Date: September 1958
Creator: Barton, David M.; Helmick, Herbert H. & Jarvis, G. A.
Partner: UNT Libraries Government Documents Department

Implementation of the RRC-KI Neutron Flux Correction Methodology in the RELAP5-3D Code

Description: The International Atomic Energy Agency has sponsored a program, “Accident Analysis and its Associated Training Programme for RBMK-1000 Kursk-1 NPP (Phase II)”. Under the auspices of this program, Reactor Research Centre “Kurchatov Institute” (RRC-KI) has implemented a Neutron Flux Correction Methodology in Version 1.2.2 of the RELAP5-3D code. The implementation was done on the RINSC workstation in Moscow, and is documented in Reference 1. Because access to the RELAP5-3D source coding by RRC-KI was limited to only the subroutines needed for the interface to the flux correction subroutine, the implementation was done using local variable arrays. The input detector data were accessed by the subroutine via local data files, residing on the computer disc storage. INEEL was then tasked with providing a permanent installation in the current release of the code. Therefore, the subject of this report is implementation of the Neutron Flux Correction Methodology in RELAP5-3D Version 2.0.3 as a permanent feature.
Date: August 1, 2003
Creator: Fisher, J. E.
Partner: UNT Libraries Government Documents Department

METHOD AND APPARATUS FOR QUANTITATIVE NEUTRON ACTIVATION ANALYSIS OF LARGE SAMPLES

Description: A method and apparatus were devised to irradiate multiple samples of large physical size simultaneously in a nonuniform neutron flux. A capsule containing the samples and flux monitors is rotated about an axis at constant speed with samples fixed in a symmetrical geometry so that each position receives the same integrated neutron flux. (auth)
Date: January 1, 1964
Creator: Hutchin, W H
Partner: UNT Libraries Government Documents Department

THERMAL FLUX PROGRAM FOR A SLAB SYSTEM

Description: A program is described for computing a quantity, Q, proportional to the neutron scalar flux, in an infinite heterogeneous slab system. The system is generated by a two-region unit cell. Q is the average track length per unit length, in a given interval, arising from the neutron traffic established by a spatially distributed monenergetic source. The program is coded for the IBM 704 computer. (auth)
Date: February 1, 1958
Creator: Beeler, J.R. & Popp, J.D.
Partner: UNT Libraries Government Documents Department

Numerical Solution of the One-Group Space-Independent Reactor Kinetics Equations for Neutron Density Given the Excess Reactivity

Description: The advantages and shortcomings of the codes currently in use at Argonne (RE-13 and RE-129) are discussed. A new method of solution, which has increased accuracy, stability for exceptionally large integration intervals, and a procedure for automatically changing the integration interval as the nature of the problem changes, is developed. (auth)
Date: February 1, 1960
Creator: Kaganove, J. J.
Partner: UNT Libraries Government Documents Department

Argonaut Automatic Flux Controller Design Report

Description: Report issued by the Argonne National Laboratory over design studies conducted on Argonaut reactors for training and research purposes. As stated in the abstract, "the design presented in the form of a steady-state analysis based upon the small-signal linearization of the reactor kinetics transfer function. The measured controller performance and construction details of the equipment is given" (p. 7). This report includes tables, illustrations, and photographs.
Date: January 1960
Creator: Gerba, A., Jr.
Partner: UNT Libraries Government Documents Department

Designing a Gas Test Loop for the Advanced Test Reactor

Description: The Generation IV Reactor Program and the Advanced Fuel Cycle Initiative are investigating some new reactor concepts which require extensive materials and fuels testing in a fast neutron spectrum. The capability to test materials and fuels in a fast neutron flux in the United States is very limited to non-existent. It has been proposed to install a gas test loop (GTL) in one of the lobes of the Advanced Test Reactor (ATR) at the Idaho National Laboratory and harden the spectrum to provide some fast neutron flux testing capabilities in the United States. This paper describes the neutronics investigation into the design of the GTL for the ATR.
Date: November 1, 2005
Creator: Parry, James R.
Partner: UNT Libraries Government Documents Department