11,584 Matching Results

Search Results

Advanced search parameters have been applied.

The current status of ARAC (Atmospheric Release Advisory Capability) and its application to the Chernobyl event

Description: The Atmospheric Release Advisory Capability (ARAC) project, developed by the Lawrence Livermore National Laboratory (LLNL), provides real-time dose assessments and estimates of the extent of surface contamination that may result from an atmospheric release of radioactivity. It utilizes advanced computer-based data communication and processing systems to acquire the meteorological and source term information needed by the three-dimensional atmospheric dispersion models to derive the consequence assessments. The ARAC responded to the recent Chernobyl reactor accident in the Soviet Union by estimating the source term and the radiation dose distribution due to exposure to the radioactive cloud over Europe and the Northern Hemisphere. This analysis revealed that approximately 50% of the estimated core inventories of I-131 and Cs-137 were released. The estimated committed effective dose equivalent due to inhalation of radioactivty during cloud passage is of the order of 10 mrem within parts of Scandinavia and eastern Europe, while most of the populations within central Europe were exposed to levels ranging from 1 to 10 mrem. The amount of Cs-137 released by the Chernobyl accident far exceeds that released by previous reactor accidents, but is only about 6% of the Cs-137 produced by the atmospheric weapon testing programs. 9 refs., 4 figs., 2 tabs.
Date: October 1, 1986
Creator: Gudiksen, P. H.; Sullivan, T. J. & Harvey, T. F.
Partner: UNT Libraries Government Documents Department

Design-related inherent safety characteristics in large LMFBR power plants

Description: Design-related safety-enhancing features such as (1) extended pump coastdown, (2) increased negative reactivity feedbacks, (3) reduced sodium void reactivity, and (4) self-actuated shutdown systems are evaluated. Primary emphasis is placed on preventing or limiting core damage. Attention is also given to features aimed at mitigation of the energetics potential of hypothetical core-disruptive accidents.
Date: January 1, 1976
Creator: Tzanos, C. P.; Barthold, W. P.; Bowers, C. H.; Ferguson, D. R.; Prohammer, F. G. & van Erp, J. B.
Partner: UNT Libraries Government Documents Department

Improved and verification of fast reactor safety analysis techniques. Annual summary, March 1, 1975--February 29, 1976

Description: Analyses of the Kiwi-TNT and SNAPTRAN-2 experiments have been performed with the VENUS-II fast-reactor disassembly code. The results show that VENUS-II provides an adequate characterization of these experiments. As is the case for LMFBRs, the excursions were initially turned over by temperature feedback effects, with ultimate shutdown coming from core disassembly. The calculated fission energies agree with the experimental values to within about 50 percent for the Kiwi excursion and 10 percent for the SNAPTRAN-2 experiment. The results of the analyses are being evaluated to determine the reasons for the remaining differences. It appears that part of the difference observed in the Kiwi-TNT analysis could relate to not explicitly treating the heat-transfer from the beaded fuel (a problem not present in LMFBR calculations). Both analyses also have uncertainties associated with the new equation-of-state that had to be added to VENUS-II to allow treatment of the core materials not used in fast reactors. Finally, there are uncertainties in the temperature feedback coefficients being used. In general, the uncertainties associated with applying VENUS-II to LMFBR excursions should be even smaller than those encountered in these experimental comparisons. This is because the temperature (Doppler) coefficients and core material equations-of-state are better known, and the complications associated with heat transfer from the beaded fuel are not present.
Date: January 1, 1976
Creator: Jackson, J. F. & Bott, T. F.
Partner: UNT Libraries Government Documents Department

HEXERE12: computer program for the transient thermal-hydraulic analysis of high temperature gas-cooled reactors

Description: HEXERE12 code development has provided a complex computer code, which is capable of analyzing High Temperature Gas-Cooled Reactors (HTGR) during accidents and providing most of the information necessary for HTGR Safety Studies. HEXERE12 is designed to solve steady-state and transient three-dimensional heat conduction, coupled with convection to axially flowing coolant in a large number of parallel channels. Both the coolant temperature and flow distribution among the channels are influenced by the heat transferred into coolant. The thermal conductivity, density, and specific heat may be spatially and temperature dependent and may further be modified by HTGR design parameters. The conductivity can be anisotropic. Heat generation rates may be dependent on time, position, and HTGR design parameters. The boundary temperatures may be time-dependent. The boundary conditions may be fixed temperatures or any combination of prescribed heat flux, forced convection, and radiation from a surface to a boundary temperature. The boundary condition parameters may be time- and/or temperature-dependent. HEXERE12 can model a coolant which can be helium, air, or a helium-air mixture.
Date: May 1, 1977
Creator: Giles, G. E.; Turner, W. D.; Childs, K. W.; DeVault, R. M. & Becker, B. R.
Partner: UNT Libraries Government Documents Department

HTGR accident initiation and progression analysis status report. Volume VII. Occupational radiation exposures from gas-borne and plateout activities

Description: As a part of the Accident Initiation and Progression Analysis (AIPA) program, calculations were performed of the occupational dose rates and man-rem exposures from gas-borne and plateout activities in a reference 3000-MW(t) HTGR plant. The study included a preliminary survey to determine the most important contributors by operation or radiation source to the man-rem exposures. This survey was followed by detailed calculations for the most important cases. Median and 95 percent-confidence-level man-rem exposures per year were obtained for the gaseous activity in the containment building, moisture monitor system, analytic instrumentation, helium regeneration system, gas waste system, and reflector-block shipping. Median and 95 percent-confidence-level man-rem exposures per operation were obtained for the main-circulator removal, steam-generator tube plugging, and steam-generator removal and replacement. For each of these cases, the contributions to the man-rem exposures were calculated for the important isotopes.
Date: January 30, 1976
Partner: UNT Libraries Government Documents Department

U. S. elevated temperature structural design standards: current status and future directions

Description: In the United States licensing of nuclear power plants requires that the owner of the plant demonstrate that the health and safety of the public is not and will not be endangered by the operation of the plant. That demonstration is a matter of public record and is subject to review and criticism, in an advisary hearing, by state and federal licensing authorities and any member of the public. National concensus structural design standards have been one of the responses to this form of power plant licensing since they effectively remove structural design rules from the arena of conflict. The resulting national standards tend to be generally applicable to all plant types and to relatively diverse operating conditions and material types. Code Case 1592 which is the elevated temperature nuclear design criteria is an example of such a national standard. Its development was the spontaneous outgrowth of the U.S. LMFBR program which demanded the best possible assurance of integrity. Being written within the framework of the ASME Boiler Code it was developed as a general standard, not just a special case for the FFTF or the CRBR Project. In the paper the development of Code Case 1592 is traced. The current and future technical content of the elevated temperature design standards for nuclear service are discussed. The relationship of Code Case 1592 to other ASME Standards and to certain U.S. industrial, governmental and regulatory standards is examined.
Date: January 1, 1976
Creator: Snow, A.
Partner: UNT Libraries Government Documents Department

Safety consequences of local initiating events in an LMFBR

Description: The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered.
Date: December 1, 1975
Creator: Crawford, R. M.; Marr, W. W.; Padilla, A. Jr. & Wang, P. Y.
Partner: UNT Libraries Government Documents Department

Diffusion of gases in solids: rare gas diffusion in solids; tritium diffusion in fission and fusion reactor metals. Final report

Description: Major results of tritium and rare gas diffusion research conducted under the contract are summarized. The materials studied were austenitic stainless steels, Zircaloy, and niobium. In all three of the metal systems investigated, tritium release rates were found to be inhibited by surface oxide films. The effective diffusion coefficients that control tritium release from surface films on Zircaloy and niobium were determined to be eight to ten orders of magnitude lower than the bulk diffusion coefficients. A rapid component of diffusion due to grain boundaries was identified in stainless steels. The grain boundary diffusion coefficient was determined to be about six orders of magnitude greater than the bulk diffusion coefficient for tritium in stainless steel. In Zircaloy clad fuel pins, the permeation rate of tritium through the cladding is rate-limited by the extremely slow diffusion rate in the surface films. Tritium diffusion rates through surface oxide films on niobium appear to be controlled by cracks in the surface films at temperatures up to 600/sup 0/C. Beyond 600/sup 0/C, the cracks appear to heal, thereby increasing the activation energy for diffusion through the oxide film. The steady-state diffusion of tritium in a fusion reactor blanket has been evaluated in order to calculate the equilibrium tritium transport rate, approximate time to equilibrium, and tritium inventory in various regions of the reactor blanket as a function of selected blanket parameters. Values for these quantities have been tabulated.
Date: September 1, 1976
Creator: Abraham, P. M.; Chandra, D.; Mintz, J. M.; Elleman, T. S. & Verghese, K.
Partner: UNT Libraries Government Documents Department

Density and shape factor of sodium aerosol. Progress report, April 1, 1976--June 30, 1976

Description: The following techniques have been implemented to improve the accuracy of aerosol centrifuge sampling and analysis systems; increased number of scanning electron microscopy photos, improved interpolation for sodium mass measurements, calculation of diameter of average volume, use of new standard sphere calibration photographs, and use of backup filter on aerosol centrifuge. Results of the first full scale trial sampling run for 1 lb. sodium fire in a 90 m/sup 3/ chamber gave density modification factors of 0.22 to 0.41.
Date: August 1, 1976
Creator: Hinds, W. & First, M. W.
Partner: UNT Libraries Government Documents Department

Experimental stress analysis for four 24-in. ANSI standard B16. 9 tees

Description: The experimental stress analysis and low cycle fatigue tests of four tees tested by Combustion Engineering, Inc. (E-E) under subcontract to Union Carbide Nuclear Division are described. These tests are part of the ORNL Design Criteria for Piping and Nozzles Program which is being conducted for the development of design criteria for nuclear power plant service piping components. The test assemblies were fabricated at C-E from commercially obtained ANSI B16.9 tees and matching diameter steel pipes welded to the tees, with suitable and closures and fixtures for applying the loads.
Date: January 1, 1976
Creator: Hayes, J. K. & Moore, S. E.
Partner: UNT Libraries Government Documents Department

Improvement and verification of fast reactor safety analysis techniques. Progress report, April 1, 1977--June 30, 1977

Description: An energy balance was made to determine the effect of the latent heat of vaporization of fuel and steel on temperature changes of the fuel and steel. Calorimeter runs were made on the reaction of DMSO with acetyl chloride to determine the effects of concentration and injection speed on the power of the reaction. Within a limited range, the power is proportional to both the concentration and injection rate. A plot of void fraction as a function of concentration was made. The concentration at which the reaction suddenly produces a large void fraction is between 3 and 4 molar.
Date: January 1, 1977
Creator: Barker, D. H. & Wheeler, P. A.
Partner: UNT Libraries Government Documents Department

LMFBR safety program. Annual technical progress report, government fiscal year 1976 and 1976T. [Sodium, fuel, and fission, product aerosol behavior]

Description: Progress is summarized in LMFBR safety studies related to the characterization of sodium fires and fission products. Included are sections on SOMIX Code development, sodium splash dispersal, aerosol leakage, aerosol model improvement, characterization of aerosol source term, large scale molten fuel tests, fuel and fission product release from burning sodium, and iodine attenuation. (DG)
Date: January 10, 1977
Partner: UNT Libraries Government Documents Department

Numerical simulation of groundwater flow and contaminant transport at the K, L, and P areas of the Savannah River Site, Aiken, South Carolina

Description: The Department of Energy (DOE) is preparing an Environmental Impact Statement (EIS) as part of the process for continuing operation of three reactors at the Savannah River Site (SRS). As required by the National Environmental Policy Act (NEPA), the EIS must address the potential environmental consequences to human health and the environment of this major federal action.'' Some of the possible consequences are related to subsurface transport of radionuclides released to seepage basins during normal reactor operation. To assist in the evaluation of the potential subsurface environmental impacts of these releases, Camp Dresser McKee Inc. (CDM) was contracted in June of 1989 to develop a three-dimensional groundwater flow and contaminant transport model which will simulate the movement of radionuclides at each of the reactor areas after they enter the groundwater system through the seepage basins. This report describes the development, calibration, and simulation results of the groundwater flow and contaminant transport model developed for this task. 10 refs., 63 figs., 11 tabs.
Date: November 1, 1989
Partner: UNT Libraries Government Documents Department

Reactor primary coolant system pipe rupture study. Progress report No. 33, January--June 1975. [BWR]

Description: The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase 1), analytical and experimental efforts (Phase 2) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, (c) studies directed at quantifying weld sensitization in Type 304 stainless steel, (d) support studies to characterize the electrochemical potential behavior of a typical BWR primary water environment and (e) special tests related to simulation of fracture surfaces characteristic of IGSCC field failures.
Date: October 1, 1975
Partner: UNT Libraries Government Documents Department

Development of a lumped parametric model for scoping investigations of uncertainties in fast reactor probabilistic safety analysis. Progress report, October 10, 1974--October 10, 1975

Description: The objective of the researh reported is to explore the possibility of the development of a novel reactor safety analysis methodology suitable for a parametric investigation of uncertainties in the progression of severe fast reactor accidents. The essential feature of this approach is a description of the reactor state by means of volumetric distributions (the distribution of volume of reactor materials, such as coolant, clad, and fuel, with temperature and in the case of fuel material, also with power). Stationary volumetric distributions are computed from detailed spatial temperature and power distributions of materials in the steady state reactor. Stationary volumetric distributions and other reactor physics quantities form the input for the reactor transient calculations in which the accident progression is described by the behavior of transient volumetric distributions. The report discusses the representation of spatial temperature distributions, the theory and calculation of stationary volumetric distributions, and includes examples of single subassembly and reactor distributions. The status of reactor neutronic code development and application is discussed and results are displayed.
Date: January 1, 1975
Creator: Ott, K. O. & Luck, L. B.
Partner: UNT Libraries Government Documents Department

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 environment.
Date: January 1, 1976
Partner: UNT Libraries Government Documents Department

Nuclear safety characterization of sodium fires and fast reactor fission products. Quarterly technical progress report, January--March 1976

Description: Progress is reported in the areas of sodium jet dispersed tests, SOMIX code development, iodine attenuation tests, aerosol leakage tests, characterization of aerosols from vaporized fuel, and high-temperature properties of fuel materials.
Date: May 15, 1976
Partner: UNT Libraries Government Documents Department

Energy sources for the future

Description: The symposium program was designed for college faculty members who are teaching or plan to teach energy courses at their educational institutions. Lectures were presented on socio-economic aspects of energy development, fusion reactors, solar energy, coal-fired power plants, nuclear power, radioactive waste disposal, and radiation hazards. A separate abstract was prepared for each of 16 of the 18 papers presented; two papers were processed earlier: Residential Energy Use Alternatives to the Year 2000, by Eric Hurst (EAPA 2:257; ERA 1:25978) and The Long-Term Prospects for Solar Energy, by W. G. Pollard (EAPA 3:1008). Fourteen of the papers are included in Energy Abstracts for Policy Analysis. (EAPA).
Date: April 1, 1977
Creator: Duggan, J. L. & Cloutier, R. J. (eds.)
Partner: UNT Libraries Government Documents Department

Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix V. Quantitative results of accident sequences. [PWR and BWR]

Description: The methods used to quantify the accident sequences previously defined in Appendix I that can potentially lead to the release of significant amounts of radioactivity from nuclear power plants are presented. The objective of this quantification was to provide the input information needed to perform calculations of the consequences of these accident sequences as described in Appendix VI.
Date: October 1, 1975
Partner: UNT Libraries Government Documents Department

Intrusion of fluid into the inflow branch of a 180/sup 0/-approach mixing tee

Description: When the flow rates in the two inlet branches of a 180/sup 0/-approach mixing tee are greatly different, it is possible that the fluid with the high velocity may intrude into the conduit in which the low velocity fluid is flowing. It is shown that such an intrusion should not extend over many pipe diameters. However, if the faster flowing fluid is also the warmer, buoyancy forces may be generated through heat transfer. This in turn may lead to density stratification in what would normally be the cooler fluid's inlet conduit. An extensive eddy develops in this branch of the tee which carries warm fluid many diameters in the upstream direction of the cooler fluid. In the laboratory such an intrusion of warm fluid in the cool fluid branch yields large temperature differences between the top and bottom of the pipe. Such behavior in prototypic systems could produce deleterious thermal stresses. Two mathematical models have been developed to estimate the extent of this density-driven intrusion. One is an inviscid model which incorporates two additional simplifying assumptions to give an initial estimate of the significance of the temperature difference and fluid velocity. This estimate is an initial step in an iterative procedure for a numerical solution scheme. The second method which is presented is a perturbation solution for a low Reynolds number flow. The matching of the solution in two regimes will require the numerical solution of equations to determine the associated coefficients. Once this is done streamline patterns can be drawn for a variety of Froude, Reynolds, and Prandtl numbers.
Date: September 1, 1976
Creator: Debler, W.
Partner: UNT Libraries Government Documents Department

KEWB facilities decontamination and disposition. Final report

Description: The decontamination and disposition of the KEWB facilities, Buildings 073, 643, 123, and 793, are complete. All of the facility equipment, including reactor enclosure, reactor vessel, fuel handling systems, controls, radioactive waste systems, exhaust systems, electrical services, and protective systems were removed from the site. Buildings 643, 123, and 793 were completely removed, including foundations. The floor and portions of the walls of Building 073 were covered over by final grading. Results of the radiological monitoring and the final survey are presented. 9 tables, 19 figures. (auth)
Date: February 25, 1976
Creator: Ureda, B. F.
Partner: UNT Libraries Government Documents Department

Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix I. Accident definition and use of event trees. [PWR and BWR]

Description: Information is presented concerning accident definition and use of event trees, event tree methodology, potential accidents covered by the reactor safety study, analysis of potential accidents involving the reactor core, and analysis of potential accidents not involving the core.
Date: October 1, 1975
Partner: UNT Libraries Government Documents Department

Resolution of geometrical configurations by a 3-D staggered mesh system

Description: Any computer program development of thermal hydraulic analyses must begin with a relationship between the grid system and the geometric configuration. The relationship chosen sets the limits of possible applications and affects virtually all programming details of the code. In the COMMIX code development, a staggered mesh system is being employed in finite differencing the governing differential equation. A methodology has been developed to extend applications of the staggered mesh system to complicated three-dimensional geometrical configurations. The methodology presented here employs x-y-z cartesian coordinates and modifies the finite differencing equations to resolve irregular three-dimensional geometrical configurations in a staggered mesh system. More important, it lends itself to a consistant approach to handle various types of boundary conditions.
Date: January 1, 1977
Creator: Domanus, H. M. & Sha, W. T.
Partner: UNT Libraries Government Documents Department