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OSCIL: one-dimensional spring-mass system simulator for seismic analysis of high temperature gas cooled reactor core

Description: OSCIL is a program to predict the effects of seismic input on a HTGR core. The present model is a one-dimensional array of blocks with appropriate spring constants, inter-elemental and ground damping, and clearances. It can be used more generally for systems of moving masses separated by nonlinear springs and dampers.
Date: January 1, 1976
Creator: Lasker, L. (ed.)
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

HTGR accident initiation and progression analysis status report. Volume VI. Event consequences and uncertainties demonstrating safety R and D importance of fission product transport mechanisms

Description: Five accident conditions are considered in an analysis of their radiological consequences. The five accident conditions are core heatup resulting from loss of offsite power and earthquake; reheater tube leak; slow depressurization; rapid depressurization; and steam ingress from steam generator main bundle tube rupture. Consequence assessments are presented in the form of radiological doses in rem to representative site boundaries of 500m and 2500m and man-rem doses to the surrounding environmen… more
Date: January 1, 1976
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

LARC-1: a Los Alamos release calculation program for fission product transport in HTGRs during the LOFC accident

Description: The theoretical and numerical data base development of the LARC-1 code is described. Four analytical models of fission product release from an HTGR core during the loss of forced circulation accident are developed. Effects of diffusion, adsorption and evaporation of the metallics and precursors are neglected in this first LARC model. Comparison of the analytic models indicates that the constant release-renormalized model is adequate to describe the processes involved. The numerical data base fo… more
Date: October 1, 1976
Creator: Carruthers, L. M. & Lee, C. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Thermal response of core and central-cavity components of a high-temperature gas-cooled reactor in the absence of forced convection coolant flow. [NATCON code]

Description: A means of determining the thermal responses of the core and the components of a high-temperature gas-cooled reactor after loss of forced coolant flow is discussed. A computer program, using a finite-difference technique, is presented together with a solution of the confined natural convection. The results obtained are reasonable and demonstrate that the computer program adequately represents the confined natural convection.
Date: September 1, 1976
Creator: Whaley, R. L. & Sanders, J. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

HEXEREI: a multi-channel heat conduction convection code for use in transient thermal hydraulic analysis of high-temperature, gas-cooled reactors. Interim report

Description: A description is given of the development and verification of a generalized coupled conduction-convection, multichannel heat transfer computer program to analyze specific safety questions involving high temperature gas-cooled reactors (HTGR). The HEXEREI code was designed to provide steady-state and transient heat transfer analysis of the HTGR active core using a basic hexagonal mesh and multichannel coolant flow. In addition, the core auxiliary cooling systems were included in the code to prov… more
Date: May 1, 1976
Creator: Giles, G. E.; DeVault, R. M.; Turner, W. D. & Becker, B. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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SIMMER-I, an LMFBR disrupted core analysis code

Description: SIMMER-I is a coupled neutronic, fluid dynamic computer program designed to analyze the dynamics of LMFBR disrupted cores. Either point kinetics or space dependent neutronics models can be used to calculate neutronic feedback effects. Multicomponent, two-phase flow models are used to predict the large scale material motion during core disruptive accidents. The effects of solid material on the fluid flow is included by simple models for core structure and frozen material. Mass, momentum, and ene… more
Date: January 1, 1976
Creator: Smith, L. L.; Boudreau, J. E.; Bell, C. R.; Bleiweis, P. B.; Barnes, J. F. & Travis, J. R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Multidomain multiphase fluid mechanics

Description: A set of multiphase field equations--conversion of mass, momentum and energy--based on multiphase mechanics is developed. Multiphase mechanics applies to mixtures of phases which are separated by interfaces and are mutually exclusive. Based on the multiphase mechanics formulation, additional terms appear in the field equations when the physical size of the dispersed phase (bubble or droplet) is many times larger than the inter-molecular spacing. These terms are the inertial coupling due to virt… more
Date: October 1, 1976
Creator: Sha, W. T. & Soo, S. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Distribution of burst pressure for tubes

Description: In a nuclear reactor, tubes are pressurized from interior by coolant, while externally no pressure is applied on them. The pressure that causes any of the tubes to burst is random and has certain distribution. By using the presently available data from stress-strain experiment, mathematical procedure for finding the distribution form of the ultimate stress is made and is justified theoretically and empirically. The distribution function obtained is important in analyzing the problem of loss of … more
Date: October 1, 1976
Creator: Kao, C S
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Computer method for analyzing HTGR core block response to seismic excitation

Description: An analytical method is developed to describe the seismic response of a High Temperature Gas-Cooled Reactor core. Models characterized by a small number of degrees-of-freedom are used to examine the large-deflection inelastic behavior of HTGR core block assemblages. The equations of motion are integrated numerically by Newmark's method, and close tolerances on the solution are enforced by an equilibrium iteration scheme. Some uses of this method are mentioned.
Date: August 1, 1976
Creator: Merson, J. L. & Bennett, J. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Fuel coolant thermal interaction project. Quarterly progress report No. 5, July 1, 1976--September 30, 1976. [LMFBR]

Description: The objective of the work reported is to experimentally and analytically study the dominant mechanisms in fuel-coolant thermal interactions which could lead to vapor explosions. The exploration of mechanisms is focused in three areas: (1) mechanisms responsible for fragmentation in molten metal droplet experiments (including assessment of the validity of the proposed Spontaneous Nucleation Mechanism), (2) thermal stress initiated fracture as a fragmentation mechanism, and (3) possible mechanism… more
Date: January 1, 1976
Creator: Todreas, N E
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Depressurization accident analyses for the Fort St. Vrain Reactor

Description: Design-basis depressurization accident analyses for the Fort St. Vrain reactor were performed using the FLODIS (Ref. 4) code. The FLODIS code models the active core, side reflector, gas annulus between the core barrel and the PCRV liner, and the PCRV cooling system. Results are presented for the Pelton circulators operating at 10,550, 8800, and 7000 rpm. Maximum temperatures of selected components are plotted as a function of time during the transient. None of the components studied exceeded th… more
Date: November 19, 1976
Creator: Paul, D. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Core heatup and fission product release from an HTGR core in an LOFC accident. [AYERM code]

Description: The AYERM code is a computer program which has been developed for the high-temperature gas-cooled reactor (HTGR) safety research program. It is a conjunction of the heat conduction code, AYER, and a set of special subroutines. This modified AYER code can predict the time-dependent release of volatile fission products from a reactor core during a hypothetical loss-of-forced-circulation (LOFC) accident. The computation scheme is based on the finite element method. The function of the AYER code is… more
Date: August 1, 1976
Creator: Cort, G. E. & Fu, J. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Fuel pin failure models and fuel-failure thresholds for core disruptive accident analysis

Description: A fuel pin failure model has been developed which uses a failure criterion based on the Larson-Miller parameter and a stress-rupture life fraction rule. The cladding is treated as an elastic-plastic solid with arbitrary strain hardening allowed. TREAT experiment analyses have been performed which show excellent agreement between measured and predicted pin failure times, failure locations, and permanent outer strain. Using this model, the effects of certain nonprototypic irradiation conditions h… more
Date: January 1, 1976
Creator: Mast, P. K. & Scott, J. H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Acoustic detection of boiling in the Sodium Loop Safety Facility in-reactor experiment P1

Description: Acoustic data were obtained from two high-temperature lithium niobate microphones on the loop background noise and transient pressure pulses during the Sodium Loop Safety Facility (SLSF) P1 in-reactor experiment. This experiment simulated an LMFBR loss-of-piping-integrity (LOPI) transient on a nineteen element, end-of-life, enriched-UO/sub 2/ fuel assembly. The microphones were exposed to liquid sodium at a distance 4.85 meters above the reactor core at temperatures between 315/sup 0/ and 590/s… more
Date: June 1, 1976
Creator: Carey, W. M.; Anderson, T. T. & Bobis, J. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Operational safety system reliability. Progress report, May 15, 1976--August 14, 1976

Description: The current research effort is roughly divided into two areas, (1) the determination of the protection system reliability and its probability distribution based on the failure distribution of the components and (2) the study of statistical characteristics of protection system sensors in order to extract the maximum information and to better anticipate off-normal conditions to promote safety and possibly avoid costly plant outages. This report summarizes the work to date in these two areas. The … more
Date: August 1, 1976
Creator: Hockenbury, R. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Fuel Coolant Thermal Interaction Project. Quarterly progress report No. 2, October 1, 1975--December 31, 1975. [LMFBR]

Description: The objective of the work reported is to experimentally and analytically study the dominant mechanisms in fuel coolant thermal interactions which could lead to vapor explosions. The exploration of mechanisms is focused in two areas: (a) mechanisms responsible for fragmentation in molten metal droplet experiments (including assessment of the validity of the proposed spontaneous nucleation mechanism); and (b) thermal stress initiated fracture as a fragmentation mechanism. Work being performed in … more
Date: March 1, 1976
Creator: Todreas, N. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Analysis of GCFBR transients without scram

Description: The analyses of four transients imposed on a Gas Cooled Fast Breeder Reactor (GCFBR) plant are presented. These calculations were carried out using the HELAP code, which is based on RELAP 3B. The transient imposed on the system is a design basis depressurization accident (DBDA) with no scram. Furthermore, the rupture size causing the depressurization was varied over a range from .6 ft/sup 2/ to zero. In the limit of no rupture the transient imposed on the system is due only to the transient beh… more
Date: February 1, 1976
Creator: Ludewig, H
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Work-energy characterization for core-disruptive accidents. [LMFBR]

Description: A consistent parametric calculation of a number of reactor types has been carried out to compare various methods of characterizing work energy from core-disruptive accidents. While the best method of characterizing damage is to do an explicit calculation of the actual energy partition, the finite expansion to the cover gas volume is a better measure of damage for parametric studies. Also, an inspection of the calculations done reveals the small fraction of total energy generated in an excursion… more
Date: January 1, 1976
Creator: Marchateere, J.; Marciniak, T.; Bratis, J.; Fuaske, H. & Jackson, J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Development of models for dynamics of subassembly can walls during core disruption

Description: Employing thin shell theory and Hamilton's principle, the equations of motion in cylindrical coordinates with theta-symmetry are very briefly derived. These equations describe the deformation of a thin, work-hardened, elastic-plastic cylindrical wall undergoing dynamic pressure or thermal loads on its surfaces. The method of differencing these equations and the solution of same is sketched. The results of a test problem comparing this approximate method with a computation by an elaborate two-di… more
Date: January 1, 1976
Creator: Blewett, P. J. & DeVault, G. P.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Analyses of LMFBR core disruption and accident phenomena using the SIMMER-I code

Description: Two types of LMFBR safety problems are analyzed using the SIMMER-I code. A whole core analysis of a 1000 MW(e) LMFBR subject to a 20 dollar/second reactivity insertion provides insight into the large scale extended material motions which result and indicates the necessity for using space-time kinetics to adequately determine the neutronic state of this system. The second problem simulates a single subassembly dry autoclave experiment. The transport behavior of the materials depends strongly on … more
Date: January 1, 1976
Creator: Bell, C. R.; Bleiweis, P. B. & Boudreau, J. E.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Fast reactor safety experiment needs and measurement requirements

Description: The usefulness of in-reactor fast reactor safety experiments depends strongly on the quality and quantity of data which can be obtained to extend knowledge of phenomena and to allow detailed comparison with theoretical models. It is imperative that the design of new safety research facilities optimizes data acquisition possibilities within the constraints of other experiment requirements. This is most important in relation to fuel and cladding motion diagnostics. The experiment needs related to… more
Date: January 1, 1976
Creator: Stevenson, M. G.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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