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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

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

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

Final LDRD report : advanced plastic scintillators for neutron detection.

Description: This report summarizes the results of a one-year, feasibility-scale LDRD project that was conducted with the goal of developing new plastic scintillators capable of pulse shape discrimination (PSD) for neutron detection. Copolymers composed of matrix materials such as poly(methyl methacrylate) (PMMA) and blocks containing trans-stilbene (tSB) as the scintillator component were prepared and tested for gamma/neutron response. Block copolymer synthesis utilizing tSBMA proved unsuccessful so random copolymers containing up to 30% tSB were prepared. These copolymers were found to function as scintillators upon exposure to gamma radiation; however, they did not exhibit PSD when exposed to a neutron source. This project, while falling short of its ultimate goal, demonstrated the possible utility of single-component, undoped plastics as scintillators for applications that do not require PSD.
Date: September 1, 2010
Creator: Vance, Andrew L.; Mascarenhas, Nicholas; O'Bryan, Greg & Mrowka, Stanley
Partner: UNT Libraries Government Documents Department

TWO-GROUP CONSTANTS FOR REACTOR MATERIALS

Description: In order to facilitate reactor design studies a compilation of calculated two-group constants averaged over the infinite-medium flux produced by a fission source was made for approximately 80 materials of interest to reactor engineers. A comparison with available experimental age measurements is included. (auth)
Date: May 1, 1958
Creator: Stanley, M.J.
Partner: UNT Libraries Government Documents Department

THEORETICAL FEEDBACK ANALYSIS IN BOILING WATER REACTORS

Description: The dynamic behavior of boiling-water reactors for small perturbations was investigated in a systematic way. General expressions for the transfer functions associated with the individual feedback mechanisms were obtained for an arbitrary flux distribution, weighting function, and steam velocity distribution. Specific forms were derived in the case of a first power flux weighting, a uniform steam velocity distribution, and a sinusoidal flux distribution with an adjustable wave length. These forms were simplified and single time-constant transfer functions were obtained. The error involved in the lumped time-constant approximation was shown to be as large as 4 db in amplitude in certain feedback mechanisms. Theoretical results were applied to the experimental power-void transfer function obtained at Ramo-Wooldridge Research Laboratory, and to the EBWR transfer function. In the former case, the agreement was found to be reasonably good, but yet more systematic experimental data were needed to reach a definite conclusion as to the validity of the proposed model, which assumes a time lag associated with steam formation and a steam perturbation speed greater than the steady-state steam velocity. In the second application, the agreement between the experimental and calculated reactor responses was proved to be better than 5 db in amplitude and 10 deg in phase, in the entire frequency range from 0.01 to 100 rad/sec. (auth)
Date: October 1, 1960
Creator: Akcasu, A.Z.
Partner: UNT Libraries Government Documents Department

TWO-DIMENSIONAL TWO-GROUP CALCULATION OF THE ARGONAUT ONE SLAB LOADING

Description: An attempt is made to set forth in systematic form the two-energy group physics calculations for the Argonaut one-slab configuration. All assumptions are stated and full calculational details given so that the procedure used may be followed. Complete point by point flux values from the PDQ programming of the problem on the IBM-704 are given. A comparison of theoretical and experimental results is included. (W.D.M.)
Date: April 1, 1960
Creator: Moon, D.P.
Partner: UNT Libraries Government Documents Department

CALCULATION OF THERMAL NEUTRON FLUXES IN PRIMARY SHIELDS

Description: A method is presented for calculating thermal neutron fluxes in the primary shields of reactor systems which eliminates reliance on mock-up experimental data. A multigroup P/sub 1/ approach is ernployed with the spatial dependence of the neutron sttenuation adjusted through use of a point source attenuation kernel for a homogeneous hydrogenous medium. Comparison of calculation with experiment is presentad. (auth)
Date: November 1, 1959
Creator: Anderson, D.C. & Shure, K.
Partner: UNT Libraries Government Documents Department