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in subjects: "Reactors 220900* -- Nuclear Reactor Technology-- Reactor Safety"

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Analysis of hypothetical LMFBR whole-core accidents in the USA

The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for accident analysis in the USA which have been developed under DOE sponsorship and summarizes some specific applications of the codes to the current gene… more

Date: January 1, 1978

Creator: Ferguson, D.R.; Deitrich, L.W.; Brown, N.W. & Waltar, A.E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Low sodium void cores

To avoid high energy releases in LMFBR TUC accidents which are accompanied by a failure to scram with a regular shutdown system, various devices have been proposed which would add negative reactivity to the core by either bringing poison material into the core or by creating negative reactivity feedbacks coming from the thermal expansion of the core. While inherent shutdown systems (ISSs) show promise for enhancing safety by adding poison to the reactor, the trigger mechanism and the geometry o… more

Date: January 1, 1978

Creator: Barthold, W. P.; Beitel, J. C.; Lam, P. S. K.; Orechwa, Y.; Su, S. F. & Turski, R. B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Improvement and Verification of Fast Reactor Safety Analysis Techniques. Progress Report, October 1, 1978--December 31, 1978

The effect of using Acetone, Benzene, or Toluene as a solvent material in a 4M, Acetyl Chloride-Dimethyl Sulfoxide reaction system was investigated. Maximum void fraction and length of time with void greater than 50% were compared. The Acetone system produced the largest maximum void and longest time above 50% void followed by Benzene and Toluene in descending order. Seventeen waxes were tested and compared to determine a material suitable to simulate reactor containment materials (e.g., steel,… more

Date: January 1, 1978

Creator: Barker, D. H. & Wiberg, D. V.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Containment building atmosphere response during severe accidents in high temperature gas-cooled reactors

Several safety evaluations for large High Temperature Gas Cooled Reactors (HTGR), using a Prestressed Concrete Reactor Vessel (PCRV) design, have concluded that Unrestricted Core Heatup Accidents (UCHA) present the most important severe accidents, resulting in the dominant source term. While the core thermohydraulic transients for such accident sequences have been presented previously, the subject of this paper is the containment building (CB) atmosphere transient, with primary emphasis on the … more

Date: January 1, 1985

Creator: Kroeger, P. G. & Chan, B. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Results of transient overpower events on breached and unbreached fuel pins

The objective of the extended overpower tests on intact pins was to determine the pin cladding breaching thresholds vis-a-vis the Plant Protection System (PPS) trip settings, typically at approx.10 to 15% overpower. These tests emphasize slow operational-type transients in light of earlier work which suggested that irradiated mixed-oxide fuel pins may be particularly vulnerable in the slow ramp-rate regime. An overview of the extended overpower test series was previously reported. More recent r… more

Date: April 1, 1986

Creator: Strain, R. V.; Tsai, H. C.; Neimark, L. A. & Aratani, K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of nonlinear fluid structure interaction transient in fast reactors

A generalized Eulerian method is described for analyzing the fluid transients and the structural response in nuclear reactors under the postulated accident conditions. The phenomena considered are the wave propagation, slug impact, sodium spillage, bubble migration, and the fluid-structure interaction. The basic equations and numerical formulation are presented in detail. Sample calculations are given to illustrate the analysis. It is shown from the results that the implicit, iterative method u… more

Date: November 1, 1978

Creator: Wang, C. Y.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Approximate inversion method for five point difference matrix equations. [LMFBR]

SIMMER, a best estimate computer program for LMFBR disrupted core analysis, is being developed at the Los Alamos Scientific Laboratory. Recent fluid dynamics methods development for SIMMER was concerned with enhancing numerical stability and reducing calculational effort. One such development considered the evaluation of the pressure change and material density change distributions during a time step. The spatially uncoupled approximate solution to be iterated was replaced with a correct spatia… more

Date: January 1, 1979

Creator: Steinke, R.B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Growth of molten core debris pools in concrete. Progress report, March 1, 1977--November 30, 1977. [LMFBR]

Experiments have been conducted using a volumetrically-heated pool with noncondensable gas injection at the boundaries to simulate the heat transfer processes taking place in molten core debris/concrete systems. Measurements of the upward, downward, and sideward heat transfer rates have been made over wide ranges of internal Rayleigh number (6.0 x 10/sup 5/ < Ra < 2.7 x 10/sup 9/), superficial gas velocity (0 < V/sub g/ < 2.5 cm/min), and pool aspect ratio (1.32 < W/L < 5.16).… more

Date: November 1, 1977

Creator: Abdel-Khalik, S I

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

LOCA simulation in the national research universal reactor program: postirradiation examination results for the third materials experiment (MT-3)

A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR) fuel bundles as part of the Loss-of-Coolant Accident (LOCA) Simulation Program. The third materials experiment (MT-3) was the sixth in the series of thermal-hydraulic and materials deformation/rutpure experiments conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. The main objective of the experiment was to evaluate ballooning and rupture during active… more

Date: April 1, 1984

Creator: Rausch, W. N.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

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

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: 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

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: 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

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: 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]

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: 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

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: Report

Partner: UNT Libraries Government Documents Department

open access

Computational features of the CACECO containment analysis code. [LMFBR]

A code, CACECO, has been written to assist in the analysis of containment situations peculiar to sodium cooled reactors. Typically, these situations involve relatively slow energy release processes and chemical reaction heat. Two examples are given to illustrate some of the code's features. These particular cases illustrate the potential for hydrogen formation in the containment building, but show that time is available to take corrective action. The code is suitable for other problems involvin… more

Date: May 29, 1975

Creator: Peak, R. D. & Stepnewski, D. D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

SIMMER-I, an LMFBR disrupted core analysis code

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: Article

Partner: UNT Libraries Government Documents Department

open access

Multidomain multiphase fluid mechanics

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: Report

Partner: UNT Libraries Government Documents Department

open access

Distribution of burst pressure for tubes

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: Article

Partner: UNT Libraries Government Documents Department

open access

Measuring the seismic response of an HTGR core model

The main objective of the tests described was to provide experimental data in order to verify the analytical models used to develop HTGR core design loads. Most of the testing was performed on a one-fifth scale full array core model subjected to uniaxial horizontal excitation. The tests initially focused on evaluating the overall core frequency response, core damping, fuel element collision forces and displacements, and in particular, the lateral support response as a function of the excitation… more

Date: July 1, 1977

Creator: Rakowski, J. E. & Olsen, B. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Computer method for analyzing HTGR core block response to seismic excitation

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: Report

Partner: UNT Libraries Government Documents Department

open access

Comparisons of finite-element code calculations to hydrostatically loaded subassembly-duct experiments

The Liquid Metal Fast Breeder Reactor (LMFBR) core structure consists of a matrix of hexagonal subassembly ducts. Evaluation of the safety aspects of the core structure requires that reliable computational procedures be available to predict the deformation response of the subassembly configuration to postulated local energy releases. Finite-element computer codes have been developed to calculate deflections and strains of a hexcan subassembly wrapper subjected to internal and external dynamic p… more

Date: January 1, 1977

Creator: Ash, J E & Marciniak, T J

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Splitting method for computing coupled hydrodynamic and structural response

A numerical method is developed for application to unsteady fluid dynamics problems, in particular to the mechanics following a sudden release of high energy. Solution of the initial compressible flow phase provides input to a power-series method for the incompressible fluid motions. The system is split into spatial and time domains leading to the convergent computation of a sequence of elliptic equations. Two sample problems are solved, the first involving an underwater explosion and the secon… more

Date: January 1, 1977

Creator: Ash, J. E.

Item Type: Article

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]

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: Report

Partner: UNT Libraries Government Documents Department

open access

Numerical and modeling techniques used in the EPIC code. [LMFBR fuel pin failure]

EPIC models fuel and coolant motion which result from internal fuel pin pressure (from fission gas or fuel vapor) and/or from the generation of sodium vapor pressures in the coolant channel subsequent to pin failure in an LMFBR. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad rip which may be of any length or which may expand with time. One-dimensional Eulerian hydrodynamics is used to model both the motion of fuel … more

Date: January 1, 1977

Creator: Pizzica, P. A. & Abramson, P. B.

Item Type: Article

Partner: UNT Libraries Government Documents Department