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

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Final report of fuel dynamics Test E7

Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release (there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the … more

Date: April 1, 1977

Creator: Doerner, R. C.; Murphy, W. F.; Stanford, G. S. & Froehle, P. H.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Outlet plenum mixing for transient overpower conditions of a one-exit nozzle LMFBR

Two types of transient tests were employed to model a one-exit nozzle LMFBR outlet plenum. Water was used as a test fluid in the simulation of constant flowrates, Transient Overpower (TOP) conditions. In the first test, simulated fuel flow was 85% and blanket flow was 15%, whereas in the second test, the fuel flow was 100%. This allowed the assessment of the mitigating effects of blanket flow upon the exit nozzle temperature transient. The flow field was clearly three-dimensional, and a less ac… more

Date: April 1, 1978

Creator: Howard, P. A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Eulerian method for large-displacement fluid-structure interaction in reactor containments

An Eulerian method for analyzing large-displacement fluid-structure interaction in reactor containments is presented. The emphasis is on the development of a generalized hydrodynamic scheme to treat the irregular cells created by the movement of the structure with respect to the fixed Eulerian coordinates. A relaxation equation is derived from the boundary condition at the fluid-structure interface for the solution of the pressure at the interface. By combining this with the Poisson equation a … more

Date: January 1, 1977

Creator: Chang, Y. W. & Wang, C. Y.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Preliminary analysis of fission gas behavior and fuel response during an LMFBR operational transient

This summary presents results obtained from a preliminary analysis of gas behavior and oxide fuel response during an LMFBR operational transient. The DiMelfi and Deitrich model is extrapolated to operational transient regimes to delineate brittle versus ductile fuel response modes. All pertinent parameters necessary for application of the DiMelfi and Deitrich model were obtained from the LIFE-3 code.

Date: January 1, 1983

Creator: Liu, Y. Y.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Eulerian algorithm for analyzing fluid-structure interaction in the fast reactor containment

The Eulerian algorithm described is capable of analyzing two-dimensionally the fluid-structure interaction in a typical LMFBR containment consisting of various structure components. The algorithm has several special features. First, it employs the Eulerian coordinates in the formulation and enables the technique to be applicable to excursions involving large material distortions. Second, it provides rigorous hydrodynamic analyses at the fluid-structure interface and at the geometrical discontin… more

Date: January 1, 1977

Creator: Wang, C. Y.; Chang, Y. W. & Fistedis, S. H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

CORCON-MOD1 preliminary evaluation and application to safety analysis of a large LMFBR plant

The CORCON-MOD1 core material-concrete interaction code, developed at the Sandia Laboratories for LWR safety analysis, was adapted for analyzing a postulated LMFBR core melt accident.

Date: June 30, 1981

Creator: Chen, K.H. & Ray, K.S.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Sandia Sodium Purification Loop (SNAPL) description and operations manual

Sandia's Sodium Purification Loop was constructed to purify sodium for fast reactor safety experiments. An oxide impurity of less than 10 parts per million is required by these in-pile experiments. Commercial, reactor grade sodium is purchased in 180 kg drums. The sodium is melted and transferred into the unit. The unit is of a loop design and purification is accomplished by ''cold trapping.'' Sodium purified in this loop has been chemically analysed at one part per million oxygen by weight. 5 … more

Date: August 1, 1985

Creator: Acton, R. U.; Weatherbee, R. L.; Smith, L. A.; Mastin, F. L. & Nowotny, K. E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Attenuation mechanisms in the transport of in-vessel radiological source term fission products in an LMFBR

Quantifying the release of radiological source term fission products from an LMFBR reactor vessel (RV) is a necessary input to the containment analysis. To estimate this initial source term value, the distribution of the fission products and actinides inside the RV, prior to release, must be known. The in-vessel source term fission product distribution and transport behavior is also essential in assessing and mitigating the plant contamination and cleanup problems which occur from any significa… more

Date: December 1, 1983

Creator: Woodley, R. E.; Nguyen, D. H. & Hunter, C. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Numerical results obtained from the three dimensional transient single phase version of the COMMIX computer code

Results from the first three applications of the three-dimensional, transient, single-phase version of the thermal-hydraulics computer code, COMMIX, are presented. The applications include an LMFBR outlet plenum, a horizontal pipe, and a 19-pin LMFBR hexagonal fuel assembly. Each case is described and the results presented. All three applications use liquid sodium as the fluid. The plenum and pipe analyses are transient, while the fuel assembly involves a steady-state analysis.

Date: October 1, 1977

Creator: Domanus, H. M.; Schmitt, R. C. & Sha, W. T.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Response of subassembly model with internals

Analytical tools have been developed and validated by controlled sets of experiments to understand the response of an accident and/or single subassembly in an LMFBR reasonably well. They have been subjected to a variety of loadings and boundary environments. Some large subassembly cluster experiments have been performed, however little analytical work has accompanied them because of the lack of suitable analytical tools. Reported are analytical approaches to: (1) development of more sophisitica… more

Date: January 1, 1977

Creator: Kennedy, J. M. & Belytschko, T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Simulation of LMFBR excursion models by means of ICECO

Test data from experiments dealing with LMFBR containment studies are compared with containment code solutions. The features of fluid spillage, which is a subject of some concern in experiments, is modeled in the ICECO code and is stressed. Several variations of spillage solution were arrived at gaining more insight to this problem and thus serve in reducing some of the uncertainties connected with fluid spillage. Other features of the analytical models are utilized and described.

Date: January 1, 1977

Creator: Marchertas, A. H.; Wang, Y. C. & Fistedis, S. H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fission product release into the primary coolant. [HTGR]

The analytic evaluation of steady state primary coolant activity is discussed. The reported calculations account for temperature dependent fuel failure in two particle types and arbitrary radioactive decay chains. A matrix operator technique implemented in the SUVIUS code is used to solve the simultaneous equations. Results are compared with General Atomic Company's published results.

Date: January 1, 1977

Creator: Apperson, C.E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of h… more

Date: January 1, 1988

Creator: Kroeger, P. G.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of LMFBR containment response to an HCDA using a multifield Eulerian code. [MICE code]

During a hypothetical core disruptive accident (HCDA), a core meltdown may cause the fuel cladding to rupture and the fuel fragments to penetrate into the sodium coolant. The heat in the molten fuel may cause the liquid sodium to boil, changing its phase. The interactions between materials are so complicated that a single-material model with homogenized material properties is not adequate. In order to analyze the above phenomena more realistically, a Multifield Implicit Continuous-Fluid Euleria… more

Date: January 1, 1977

Creator: Chu, H.Y. & Chang, Y.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Cladding failure by local plastic instability

Cladding failure is one of the major considerations in analysis of fuel-pin behavior during hypothetical accident transients since time, location, and nature of failure govern the early postfailure material motion and reactivity feedback. Out-of-pile thermal transient tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture and that the extent of local instability limits the initial rip length. To investigate the det… more

Date: December 1, 1977

Creator: Kramer, J. M. & Deitrich, L. W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Sodemme: natural circulation thermal-hydraulics code for HTGR transient analysis

The SODEMME code provides a Solution to the One Dimensional Energy, Mass and Momentum Equations for a system of ''volumes'' and ''segments'' that comprise a circulating system containing a gaseous coolant. Input consists of a description of the system's geometry, the user's choice of some control variables, specification of the initial and time-dependent boundary conditions and setting various options for items such as pumping head, heat transfe… more

Date: July 1, 1977

Creator: Epel, L.G.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Scoping systems analysis of a 350 MWt modular liquid metal cooled reactor

The systems analysis code SASSYS was used to explore the sensitivity of the system response to various inherent reactivity feedback mechanisms and design features for a small, liquid-metal-cooled reactor during the first 1000 s following the initiation of an unprotected loss-of-flow and/or loss-of-primary-heat-removal transient. The results show that to maximize the inherent safety of small, liquid-metal-cooled reactors, inherent feedback mechanisms should be accounted for in establishing desig… more

Date: January 1, 1985

Creator: Morris, E. E.; Rhow, S. K. & Switick, D. M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Diversion analysis and safeguards measures for liquid metal fast breeder reactors

The general objective of the study is to perform a diversion analysis and an assessment of the available safeguards methods and systems for verifying inventory and flow of nuclear material in accessible and inaccessible areas of liquid-metal fast breeder reactor, LMFBR, systems. The study focuses primarily on the assembly-handling operations, assembly storage facilities, and reactor operations facilities relating to existing and/or near-term planned experimental, demonstration and prototypal re… more

Date: October 1, 1981

Creator: Persiani, P.J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Analysis of reactor material experiments investigating corium crust stability and heat transfer in jet impingement flow

Presented is an analysis of the results of the CSTI-1, CSTI-3, and CWTI-11 reactor material experiments in which a jet of molten corium initially at 3080/sup 0/K was directed downward upon a stainless steel plate. The experiments are a continuation of a program of reactor material tests investigating LWR severe accident phenomena. Objective of the present analysis is to determine the existence or nonexistence of a corium crust during impingement from comparison of the measured heatup of the pla… more

Date: January 1, 1985

Creator: Sienicki, J. J. & Spencer, B. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Structural response of large LMFBR head closures to hypothetical core disruptive accidents

A preliminary study of the dynamic response of a large (1200 MWe) pool-type deck structure to an energetic core disassembly accident was performed. The analysis proceeded as two independent, decoupled calculations: containment calculations and deck response calculations. The containment calculations were performed mainly to derive a loading pressure history for the deck structure. In part of the study, the deck was considered rigid so the derived loading pressures were conservative. In the seco… more

Date: January 1, 1977

Creator: Kulak, R.F.; Marciniak, T.J. & Belytschko, T.B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Hydrogen mixing study (HMS) in LWR type containments

A numerical technique has been developed for calculating the full three-dimensional time-dependent Navier-Stokes equations with multiple speies transport. The method is a modified form of the Implicit Continuous-fluid Eulerian (ICE) technique to solve the governing equations for low Mach number flows where pressure waves and local variations in compression and expansion are not significant. Large density variations, due to thermal and species concentration gradients, are accounted for without t… more

Date: January 1, 1983

Creator: Travis, J.R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analyses of hypothetical FCI's in a fast reactor

Parametric analyses using the SIMMER code were performed to evaluate the potential for a severe recriticality from a pressure-driven recompaction caused by an energetic FCI during the transition phase of a hypothetical accident in a fast reactor. For realistic and reasonable estimates for the assumed accident conditions, a severe recriticality was not predicted. The conditions under which a severe recriticality would be obtained or averted were identified. 10 figures, 2 tables.

Date: January 1, 1981

Creator: Padilla Jr., A.; Martin, F. J. & Niccoli, L. G.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Special consideration for the fire protection of liquid metal fast breeder reactors

In addition to the possible radiation hazard in LMFBRs, there exists the hazard of reactive metals. The principal-fire hazard associated with sodium is its rapid reaction with water, producing hydrogen. The oxygen atmosphere can be inerted or reduced, metal sinks used, combustible materials avoided, and automatic sprinklers excluded in areas where a hazard of metal coolant fires exist. Automatic fire suppression systems are discussed. Passive fire protection features are described. Liners and c… more

Date: November 1, 1983

Creator: Sansone, F.M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Comparative transient analysis of metal and oxide fueled LMFBR. [Loop-type and pool-type designs]

The neutronics and thermodynamics of an LMFBR primary system have been simulated using the DSNP simulation language. A detailed fuel pin and its subchannel model were developed and included in the DSNP library. This permits the reactor core to be simulated with any number of pins having any number of radial and axial nodes. The metal fueled core transients were compared to transients of an oxide fueled core, the conclusion being that for the same perturbations the temperature transients are fas… more

Date: September 1, 1979

Creator: Saphier, D. & Madell, J.T.

Item Type: Report

Partner: UNT Libraries Government Documents Department