32 Matching Results

Search Results

Advanced search parameters have been applied.

FREQUENCY RESPONSE ANALYSIS OF THE HRT FEEDWATER CONTROL SYSTEM

Description: A frequency response analysis is made of the HRT fuel feedwater system. The characteristics of the components are determined and used in the differential equation of the system. Using frequency response techniques, theoretical calculations of static and dynamic response characteristics are made for reactor operation at 5 Mw. The conclusions arrived at as a result of this analysis include suggestions for improving the present operation. (auth)
Date: April 1, 1958
Creator: Ball, S.J.
Partner: UNT Libraries Government Documents Department

Dynamic model verification studies for the thermal response of the Fort St. Vrain HTGR Core

Description: The safety research program for high-temperature gas-cooled reactors at ORNL is directed primarily at addressing licensing questions on the Fort St. Vrain reactor near Denver, CO. An important part of the program is to make use of experimental data from the reactor to at least partially verify the dynamic simulations that are used to predict the effects of postulated accident sequences. Comparisons were made of predictions with data from four different reactor scram (trip) events from operating power levels between 30 and 50%. An optimization program was used to rationalize the differences between predictions and measurements, and, in general, excellent agreement can be obtained by adjustment of models and parameters within their uncertainty ranges. Although the optimized models are not necessarily unique, results of the study have identified areas in which some of the models were deficient.
Date: January 1, 1980
Creator: Ball, S.J.
Partner: UNT Libraries Government Documents Department

A SIMULATION OF THE EGCR STEAM GENERATOR

Description: An analog model of the EGCR steam generator was developed and operated on the ORNL analog computer as part of a program to simulate the operation and control of the EGCR reactor plant. Equilibrium operation and the transient response of the steam generator unit to system perturbations were studied. A simultaneous solution of the basic heat transfer equations representing the performance of the unit was obtained. The model was operated initially at steady- state conditions, and then perturbations were made to gas flow, gas inlet temperature, and steam throttle valve position. The response characteristics of the model during the transients were recorded. The steam generator gas outlet temperature showed a marked degree of insensitivity to changes in gas inlet temperature. The effect of gas flow changes on gas exit temperature was slightly more pronounced. The transient behavio-r of the unit was reasonable, and the model developed indicated satisfactory operation within the design range of 20 to l00% of full power. (auth)
Date: October 1, 1961
Creator: Yarosh, M.M. & Ball, S.J.
Partner: UNT Libraries Government Documents Department

Modular high-temperature gas-cooled reactor simulation using parallel processors

Description: The MHPP (Modular HTGR Parallel Processor) code has been developed to simulate modular high-temperature gas-cooled reactor (MHTGR) transients and accidents. MHPP incorporates a very detailed model for predicting the dynamics of the reactor core, vessel, and cooling systems over a wide variety of scenarios ranging from expected transients to very-low-probability severe accidents. The simulation routines, which had originally been developed entirely as serial code, were readily adapted to parallel processing Fortran. The resulting parallelized simulation speed was enhanced significantly. Workstation interfaces are being developed to provide for user (''operator'') interaction. The benefits realized by adapting previous MHTGR codes to run on a parallel processor are discussed, along with results of typical accident analyses. 3 refs., 3 figs.
Date: January 1, 1989
Creator: Ball, S.J. & Conklin, J.C.
Partner: UNT Libraries Government Documents Department

Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

Description: One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600/sup 0/C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth.
Date: January 1, 1985
Creator: Harrington, R.M. & Ball, S.J.
Partner: UNT Libraries Government Documents Department

Performance evaluation of a selected three-ton air-to-air heat pump in the heating mode

Description: An air-to-air split system residential heat pump of nominal three-ton capacity was instrumented and tested in the heating mode under laboratory conditions. This was the second of a planned series of experiments to obtain a data base of system and component performance for heat pumps. The system was evaluated under both steady-state and frosting-defrosting conditions; sensitivity of the system performance to variations in the refrigerant charge was measured. From the steady-state tests, the heating capacity and coefficient of performance were computed, and evaluations were made of the performance parameters of the fan and fan motor units, the heat exchangers and refrigerant metering device, and the compressor. System heat losses were analyzed. The frosting-defrosting tests allowed the observation of system and component performance under dynamic conditions, and measurement of performance degradation under frosting conditions.
Date: January 1, 1980
Creator: Domingorena, A.A. & Ball, S.J.
Partner: UNT Libraries Government Documents Department

Modular high-temperature gas-cooled reactor core heatup accident simulations

Description: The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs.
Date: January 1, 1989
Creator: Ball, S.J. & Conklin, J.C.
Partner: UNT Libraries Government Documents Department

Analog Computer Study of the MSR-ORR in-Pile Pressurized Water Loop No. 1

Description: A study of the dynamic behavior of the Merchant Ship Reactor Pressurized Water Loop was made using the Reactor Controls Analog Facility. Computer curves show the predicted response of the loop temperatures to normal load changes and component failure accidents. Except for complete flow stoppage, which was not investigated here, the safety system was shown to be adequate in curbing loop temperature excursions due to postulated accidents. (auth)
Date: May 1, 1960
Creator: Ball, S. J.
Partner: UNT Libraries Government Documents Department

HTGR severe accident sequence analysis

Description: Thermal-hydraulic, fission product transport, and atmospheric dispersion calculations are presented for hypothetical severe accident release paths at the Fort St. Vrain (FSV) high temperature gas cooled reactor (HTGR). Off-site radiation exposures are calculated for assumed release of 100% of the 24 hour post-shutdown core xenon and krypton inventory and 5.5% of the iodine inventory. The results show conditions under which dose avoidance measures would be desirable and demonstrate the importance of specific release characteristics such as effective release height. 7 tables.
Date: January 1, 1982
Creator: Harrington, R.M.; Ball, S.J. & Kornegay, F.C.
Partner: UNT Libraries Government Documents Department

Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

Description: This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed.
Date: November 1, 1985
Creator: Harrington, R.M.; Ball, S.J. & Cleveland, J.C.
Partner: UNT Libraries Government Documents Department

ORNL's NRC-sponsored HTGR safety and licensing analysis activities for Fort St. Vrain and advanced reactors

Description: The ORNL safety analysis program for the HTGR was established in 1974 to provide technical assistance to the USNRC on licensing questions for both Fort St. Vrain and advanced plant concepts. The emphasis has been on development of major component and system dynamic simulation codes, and use of these codes to analyze specific licensing-related scenarios. The program has also emphasized code verification, using Fort St. Vrain data where applicable, and comparing results with industry-generated codes. By the use of model and parameter adjustment routines, safety-significant uncertainties have been identified. A major part of the analysis work has been done for the Fort St. Vrain HTGR, and has included analyses of FSAR accident scenario re-evaluations, the core block oscillation problem, core support thermal stress questions, technical specification upgrade review, and TMI action plan applicability studies. The large, 2240-MW(t) cogeneration lead plant design was analyzed in a multi-laboratory cooperative effort to estimate fission product source terms from postulated severe accidents.
Date: January 1, 1985
Creator: Ball, S.J.; Cleveland, J.C. & Harrington, R.M.
Partner: UNT Libraries Government Documents Department

Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

Description: The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes.
Date: April 4, 1978
Creator: Ball, S.J.; Cleveland, J.C. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Interactive simulations of gas-turbine modular HTGR transients and heatup accidents

Description: An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design 350-MW(t) MHTGR. Subsequently, the code was modified under DOE sponsorship to simulate the 450-MW(t) Gas Turbine (GT) design and to aid in development and design studies. Features of the code (MORECA-GT) include detailed modeling of 3-D core thermal-hydraulics, interactive workstation capabilities that allow user/analyst or ``operator`` involvement in accident scenarios, and options for studying anticipated transients without scram (ATWS) events. In addition to the detailed models for the core, MORECA includes models for the vessel, Shutdown Cooling System (SCS), and Reactor Cavity Cooling System (RCCS), and core point kinetics to accommodate ATWS events. The balance of plant (BOP) is currently not modeled. The interactive workstation features include options for on-line parameter plots and 3-D graphic temperature profiling. The studies to date show that the proposed MHTGR designs are very robust and can generally withstand the consequences of even the extremely low probability postulated accidents with little or no damage to the reactor`s fuel or metallic components.
Date: June 1, 1994
Creator: Ball, S. J. & Nypaver, D. J.
Partner: UNT Libraries Government Documents Department

MORECA-GT: Interactive simulator for gas-turbine modular HTGR transients and heatup accidents with ATWS options

Description: An interactive simulation code for studying postulated heatup accidents in modular high-temperature gas-cooled reactors (MHTGRs) has been adapted to assist with parametric design studies of the US Department of Energy`s (DOE`s) direct-cycle gas-turbine MHTGR concept. The studies show that the proposed MHTGR designs are very robust and can generally withstand the consequences of extremely low probability accidents with little or no damage to the reactor`s fuel or metallic components.
Date: March 1, 1994
Creator: Ball, S. J. & Nypaver, D. J.
Partner: UNT Libraries Government Documents Department

Simulation of the response of the Fort St. Vrain high temperature gas cooled reactor system to a postulated rod withdrawal accident

Description: Transients resulting from the accidental withdrawal of a control rod pair from the Fort St. Vrain reactor core from 100% power conditions have been analyzed with the ORTAP nuclear steam supply system simulation. This analysis was done as part of an ongoing effort to obtain an independent assessment of the HTGR system response to several postulated accidents. Results are presented and discussed.
Date: January 1, 1977
Creator: Cleveland, J.C.; Hedrick, R.A.; Ball, S.J.; Delene, J.G. & Conklin, J.C.
Partner: UNT Libraries Government Documents Department

Metal burning in graphite-moderated reactors

Description: Pinto beans, sweet corn, and zucchini squash (Cucurbita pepo var. black beauty) were grown in a randomized complete-block field/pot experiment at a site that contained the highest observed levels of surface gross gamma radioactivity within Los Alamos Canyon (LAC) at Los Alamos National Laboratory. Soils as well as washed edible and nonedible crop tissues were analyzed for various radionuclides and heavy metals. Most radionuclides, with the exception of {sup 3}H and {sup tot}U, in soil from LAC were detected in significantly higher concentrations (p <0.01) than in soil collected from regional background (RBG) locations. Similarly, most radionuclides in edible crop portions of beans, squash, and corn were detected in significantly higher (p <0.01 and 0.05) concentrations than RBG. Most soil-to-plant concentration ratios for radionuclides in edible and nonedible crop tissues from LAC were within the default values given by the Nuclear Regulatory Commission and Environmental Protection Agency. All heavy metals in soils, as well as edible and nonedible crop tissues grown in soils from LAC, were within RBG concentrations. Overall, the total maximum net positive committed effective dose equivalent (CEDE)--the CEDE plus two sigma for each radioisotope minus background and then all positive doses summed--to a hypothetical 50-year resident that ingested 160 kg of beans, corn, and squash in equal proportions, was 74 mrem y{sup -1}. This dose was below the International Commission on Radiological Protection permissible dose limit (PDL) of 100 mrem y{sup -1} from all pathways; however, the addition of other internal and external exposure route factors may increase the overall dose over the PDL. Also, the risk of an excess cancer fatality, based on 74 mrem y{sup -1}, was 3.7 x 10{sup -5} (37 in a million), which is above the Environmental Protection Agency`s (acceptable) guideline of one in a million. 25 refs.
Date: May 1, 1997
Creator: Wichner, R.P.; Ball, S.J.; Daw, C.S. & Thomas, J.F.
Partner: UNT Libraries Government Documents Department

Future prospects of baryon istability search in p-decay and n n(bar) oscillation experiments

Description: These proceedings contain thirty-one papers which review both the theoretical and the experimental status and near future of baryon instability research. Baryon instability is investigated from the vantage point of supersymmetric and unified theories. The interplay between baryogenesis and antimatter is examined. Double beta decay experiments are discussed. The huge Icarus experiment is described with its proton decay capabilities. Neutron-antineutron oscillations investigations are presented, especially efforts with ultra-cold neutrons. Individual papers are indexed separately on the Energy Data Base.
Date: November 1, 1996
Creator: Ball, S.J. & Kamyshkov, Y.A.
Partner: UNT Libraries Government Documents Department

GRSAC Users Manual

Description: An interactive workstation-based simulation code (GRSAC) for studying postulated severe accidents in gas-cooled reactors has been developed to accommodate user-generated input with ''smart front-end'' checking. Code features includes on- and off-line plotting, on-line help and documentation, and an automated sensitivity study option. The code and its predecessors have been validated using comparisons with a variety of experimental data and similar codes. GRSAC model features include a three-dimensional representation of the core thermal hydraulics, and optional ATWS (anticipated transients without scram) capabilities. The user manual includes a detailed description of the code features, and includes four case studies which guide the user through four different examples of the major uses of GRSAC: an accident case; an initial conditions setup and run; a sensitivity study; and the setup of a new reactor model.
Date: February 1, 1999
Creator: Ball, S.J. & Nypaver, D.J.
Partner: UNT Libraries Government Documents Department

Automated operator procedure prompting for startup of Experimental Breeder Reactor-2

Description: This report describes the development of an operator procedure prompting aid for startup of a nuclear reactor. This operator aid is a preliminary design for a similar aid that eventually will be used with the Advanced Liquid Metal Reactor (ALMR) presently in the design stage. Two approaches were used to develop this operator procedure prompting aid. One method uses an expert system software shell, and the other method uses database software. The preliminary requirements strongly pointed toward features traditionally associated with both database and expert systems software. Database software usually provides data manipulation flexibility and user interface tools, and expert systems tools offer sophisticated data representation and reasoning capabilities. Both methods, including software and associated hardware, are described in this report. Proposals for future enhancements to improve the expert system approach to procedure prompting and for developing other operator aids are also offered. 25 refs., 14 figs.
Date: November 1, 1990
Creator: Renshaw, A.W.; Ball, S.J. & Ford, C.E.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor safety studies for the Division of Accident Evaluation. Quarterly progress report, April 1-June 30, 1985

Description: Modeling, code development, and analyses of the modular High-Temperature Gas-Cooled Reactor (HTGR) continued with work on the side-by-side design. Fission-product release and transport experiments were completed. Sections of an HTGR safety handbook were written.
Date: February 1, 1986
Creator: Ball, S.J.; Cleveland, J.C.; Harrington, R.M. & Wilson, J.H.
Partner: UNT Libraries Government Documents Department