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Reflected kinetics model for nuclear space reactor kinetics and control scoping calculations

Description: The objective of this research is to develop a model that offers an alternative to the point kinetics (PK) modelling approach in the analysis of space reactor kinetics and control studies. Modelling effort will focus on the explicit treatment of control drums as reactivity input devices so that the transition to automatic control can be smoothly done. The proposed model is developed for the specific integration of automatic control and the solution of the servo mechanism problem. The integration of the kinetics model with an automatic controller will provide a useful tool for performing space reactor scoping studies for different designs and configurations. Such a tool should prove to be invaluable in the design phase of a space nuclear system from the point of view of kinetics and control limitations.
Date: May 1, 1986
Creator: Washington, K.E.
Partner: UNT Libraries Government Documents Department

Cooperative business management strategies for the U.S. integrated textile complex

Description: The mission of the American Textile (AMTEX{trademark}) Partnership is to engage the unique technical resources of the Department of Energy National Laboratories to work with the US Integrated Textile Complex (US ITC) and research universities to develop and deploy technologies that will increase the competitiveness of the US ITC. The objectives of the Demand Activated Manufacturing Architecture (DAMA) project of AMTEX are: (1) to determine strategic business structure changes for the US ITC; (2) to establish a textile industry electronic marketplace, (3) to provide methods for US ITC education ad implementation of an electronic marketplace. The Enterprise Modeling and Simulation Task of DAMA is focusing on the first DAMA goal as described in another paper of this conference. The Cooperative Business Management (CBM) Task of DAMA is developing computer-based tools that will render system-wide information accessible for improved decision making. Three CBM strategies and the associated computer tools being developed to support their implementation are described in this paper. This effort is addressing the second DAMA goal to establish a textile industry electronic marketplace in concert with the Connectivity and Infrastructure Task of DAMA. As the CBM tools mature, they will be commercialized through the DAMA Education, Outreach and Commercialization Task of DAMA to achieve the third and final DAMA goal.
Date: December 31, 1995
Creator: Washington, K.E.
Partner: UNT Libraries Government Documents Department

Direct containment heating models in the CONTAIN code

Description: The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.
Date: August 1, 1995
Creator: Washington, K.E. & Williams, D.C.
Partner: UNT Libraries Government Documents Department

CONTAIN code analyses of direct containment heating (DCH) experiments

Description: In some nuclear reactor core melt accidents, a potential exists for molten core debris to be dispersed into the containment under high pressure. Resulting energy transfer to the containment atmosphere can pressurize the containment. This process, known as direct containment heating (DCH), has been the subject of extensive experimental and analytical programs sponsored by the US Nuclear Regulatory Commission (NRC). DCH modeling has been a major focus for the development of the CONTAIN code. In support of the peer review, extensive analyses of DCH experiments were performed in order to assess the CONTAIN code`s DCH models and improve understanding of DCH phenomenology. The present paper summarizes this assessment effort.
Date: June 1, 1995
Creator: Williams, D.C.; Griffith, R.O.; Tadios, E.L. & Washington, K.E.
Partner: UNT Libraries Government Documents Department

User`s guide for the KBERT 1.0 code: For the knowledge-based estimation of hazards of radioactive material releases from DOE nuclear facilities

Description: The possibility of worker exposure to radioactive materials during accidents at nuclear facilities is a principal concern of the DOE. The KBERT software has been developed at Sandia National Laboratories under DOE support to address this issue by assisting in the estimation of risks posed by accidents at chemical and nuclear facilities. KBERT is an acronym for Knowledge-Based system for Estimating hazards of Radioactive material release Transients. The current prototype version of KBERT focuses on calculation of doses and consequences to in-facility workers due to accidental releases of radioactivity. This report gives detailed instructions on how a user who is familiar with the design, layout and potential hazards of a facility can use KBERT to assess the risks to workers in that facility. KBERT is a tool that allows a user to simulate possible accidents and observe the predicted consequences. Potential applications of KBERT include the evaluation of the efficacy of evacuation practices, worker shielding, personal protection equipment and the containment of hazardous materials.
Date: July 1, 1995
Creator: Browitt, D.S.; Washington, K.E. & Powers, D.A.
Partner: UNT Libraries Government Documents Department

CONTAIN code analyses of direct containment heating (DCH) experiments: Model assessment and phenomenological interpretation

Description: Models for direct containment heating (DCH) in the CONTAIN code for severe accident analysis have been reviewed and a standard input prescription for their use has been defined. The code has been exercised against a large subset of the available DCH data base. Generally good agreement with the experimental results for containment pressurization ({Delta}P) and hydrogen generation has been obtained. Extensive sensitivity studies have been performed which permit assessment of many of the strengths and weaknesses of specific model features. These include models for debris transport and trapping, DCH heat transfer and chemistry, atmosphere-structure heat transfer, interactions between nonairborne debris and blowdown steam, potential effects of debris-water interactions, and hydrogen combustion under DCH conditions. Containment compartmentalization is an important DCH mitigator in the calculations, in agreement with experimental results. The CONTAIN model includes partially parametric treatments for some processes that are not well understood. The importance of the associated uncertainties depends upon the details of the DCH scenario being analyzed. Recommended sensitivity studies are summarized that allow the user to obtain a reasonable estimate of the uncertainties in the calculated results.
Date: May 12, 1995
Creator: Williams, D.C.; Griffith, R.O.; Tadios, E.L. & Washington, K.E.
Partner: UNT Libraries Government Documents Department

Modeling direct containment heating phenomena with CONTAIN 1. 12

Description: CONTAIN is a detailed mechanistic computer code developed at Sandia National Laboratories for the integrated analysis of light water reactor severe accident containment phenomena. The most recent version of the code, CONTAIN 1.12, incorporates models for the phenomena of high pressure melt ejection (HPME) and the subsequent processes collectively known as Direct Containment Heating (DCH). CONTAIN 1.12 was used to model the Limited Flight Path 8A (LFP8A) experiment conducted at the Surtsey test facility at Sandia National Laboratories. In the experiment, 50 kg of molten thermite was injected into a scale model of the Surry cavity and then blown into the Surtsey vessel by high pressure steam. A seven-cell best-estimate CONTAIN model, using only a minimum of measured data, was used to simulate the LFP8A experiment. A comparison of the experimental and calculated results indicated that CONTAIN 1.12 was accurately modeling the physical processes involved in DCH phenomena, but the method of injecting the molten debris into the cavity in the CONTAIN model was causing the code to overpredict the chemical reaction and heat transfer rates between the molten debris and the system atmosphere. CONTAIN 1.12 predicted the peak vessel pressure to within less than 2% of the experimental value, but missed the timing on the pressure peak by approximately 1.75 s over the course of a 10 s calculation. 6 refs., 6 figs.
Date: January 1, 1991
Creator: Griffith, R.O.; Russell, N.A. & Washington, K.E.
Partner: UNT Libraries Government Documents Department

A simplified model of aerosol removal by natural processes in reactor containments

Description: Simplified formulae are developed for estimating the aerosol decontamination that can be achieved by natural processes in the containments of pressurized water reactors and in the drywells of boiling water reactors under severe accident conditions. These simplified formulae were derived by correlation of results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior under accident conditions. Monte Carlo uncertainty analyses of decontamination by natural aerosol processes are reported for 1,000, 2,000, 3,000, and 4,000 MW(th) pressurized water reactors and for 1,500, 2,500, and 3,500 MW(th) boiling water reactors. Uncertainty distributions for the decontamination factors and decontamination coefficients as functions of time were developed in the Monte Carlo analyses by considering uncertainties in aerosol processes, material properties, reactor geometry and severe accident progression. Phenomenological uncertainties examined in this work included uncertainties in aerosol coagulation by gravitational collision, Brownian diffusion, turbulent diffusion and turbulent inertia. Uncertainties in aerosol deposition by gravitational settling, thermophoresis, diffusiophoresis, and turbulent diffusion were examined. Electrostatic charging of aerosol particles in severe accidents is discussed. Such charging could affect both the coagulation and deposition of aerosol particles. Electrostatic effects are not considered in most available models of aerosol behavior during severe accidents and cause uncertainties in predicted natural decontamination processes that could not be taken in to account in this work. Median (50%), 90 and 10% values of the uncertainty distributions for effective decontamination coefficients were correlated with time and reactor thermal power. These correlations constitute a simplified model that can be used to estimate the decontamination by natural aerosol processes at 3 levels of conservatism. Applications of the model are described.
Date: July 1, 1996
Creator: Powers, D.A.; Washington, K.E.; Sprung, J.L. & Burson, S.B.
Partner: UNT Libraries Government Documents Department

User's guide for the KBERT 2.0 code

Description: The possibility of worker exposure to radioactive materials during accidents at nuclear facilities is a principal concern of the DOE. The KBERT analysis tool has been developed at Sandia National Laboratories under DOE support to address this issue by assisting in the estimation of risks posed by accidents at chemical and nuclear facilities. KBERT is an acronym for Knowledge-Based system for Estimating hazards of Radioactive material release Transients. KBERT's primary purpose is to predict doses to in-facility workers due to accidental releases of radioactivity. Models are also in KBERT for predicting doses to the public based upon plume dispersal models. This report gives detailed instructions on how a user, starting with knowledge of design, layout and potential hazards of a facility, can use KBERT to assess the risks to workers in that facility and to the public as a result of releases from the facility. A key feature of KBERT is the inclusion of the non-facility-specific material release, radioactive decay, and dose databases (i.e., knowledge bases) that might also be needed for such an assessment. The material release characteristics are based on the 1994 DOE Handbook for airborne release fractions/rates and respirable fractions for nonreactor nuclear facilities. Another important feature of KBERTis the inclusion of a transparent interface between KBERTand the Nuclear Regulatory Commission's CONTAIN code. This interface enables KBERT to use the validated and proven flow models in CONTAIN to predict inter-room airflows. Potential applications of KBERT include the evaluation of the consequences of evacuation practices, the effect of personal protection equipment, and the degree of containment of hazardous materials.
Date: May 1, 2000
Creator: Washington, K. E.; Murata, K. K.; Browitt, D. S.; Brockmann, J. E.; Griffith, R. O.; Gelbard, F. et al.
Partner: UNT Libraries Government Documents Department

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

Description: The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.
Date: December 1, 1997
Creator: Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J. et al.
Partner: UNT Libraries Government Documents Department