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Liquid Metal Reactor Program: JASPER USDOE/PNC Shielding Research Program: Technical progress report, August 1-September 30, 1986

Description: This report details activities on the JASPER Shielding Program for the time period of August 1, 1986 through September 30, 1986. This report contains the measurements in phases VI and VII, a graphite benchmark study and an alternate loop type shield design study for the Liquid Metal Reactor (LMR), respectively. This report also includes the results of analyses for phases I, II, III, V, and VI.
Date: December 31, 1986
Creator: Ingersoll, D.T.; Engle, W.W. Jr.; Muckenthaler, F.J. & Slater, C.O.
Partner: UNT Libraries Government Documents Department

Research on the climatic effects of nuclear winter: Final report

Description: The National Center for Atmospheric Research (NCAR) has undertaken a series of research efforts to develop and implement improvements to the Community Climate Model (CCM) needed to make the model more applicable to studies of the climatic effects of nuclear war. The development of the model improvements has reached a stage where implementation may proceed, and several of the developed routines are being incorporated into the next approved version of the CCM (CCM1). Formal documentation is being completed describing the specific model improvements that have been successfully implemented. This final report includes the series of annual proposals and progress reports that have guided the project.
Date: December 3, 1986
Creator: Dickinson, R.E.
Partner: UNT Libraries Government Documents Department

A Study to Determine the Biological Feasibility of a New Fish Tagging System: Annual Report, 1985-1986.

Description: An ongoing cooperative project between the Bonneville Power Administration and the National Marine Fisheries Service was initiated in 1983 to evaluate the technical and biological feasibility of adapting a new identification system to salmonids. The system is based upon the passive integrated transponder (PIT) tag. This report discusses the work completed in 1985 and is divided into laboratory and field studies. All studies were conducted with the tag implanted into the body cavity of the test fish via a 12-gauge hypodermic needle.
Date: December 1, 1986
Creator: Prentice, Earl F.; Park, D.L.; Flagg, T.A. & McCutcheon, S.
Partner: UNT Libraries Government Documents Department

COBRA-SFS (Spent-Fuel Storage) thermal-hydraulic analyses of the CASTOR-1C and REA 2023 BWR storage casks containing consolidated spent fuel

Description: Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates.
Date: December 1, 1986
Creator: Rector, D.R.; Cuta, J.M. & Lombardo, N.J.
Partner: UNT Libraries Government Documents Department

Comparison of HYDRA predictions to temperature data from two single-assembly spent fuel heat transfer tests

Description: The HYDRA computer code was used to simulate the thermal performance of an actual and a model spent fuel assembly. The HYDRA-predicted temperatures were then compared with measured data from two single-assembly test sections. The objective of this effort was to further verify the predictive capabilities of the HYDRA code for use in assessments of the hydrothermal performance of spent fuel dry storage systems. After HYDRA has been adequately evaluated and validated, the code will be documented to permit design and licensing safety analyses.
Date: December 1, 1986
Creator: McCann, R.A.
Partner: UNT Libraries Government Documents Department

Rod consolidation at the West Valley Demonstration Project

Description: A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab.
Date: December 1, 1986
Creator: Bailey, W.J.
Partner: UNT Libraries Government Documents Department

MONITOR: A computer model for estimating the costs of an integral monitored retrievable storage facility

Description: The MONITOR model is a FORTRAN 77 based computer code that provides parametric life-cycle cost estimates for a monitored retrievable storage (MRS) facility. MONITOR is very flexible in that it can estimate the costs of an MRS facility operating under almost any conceivable nuclear waste logistics scenario. The model can also accommodate input data of varying degrees of complexity and detail (ranging from very simple to more complex) which makes it ideal for use in the MRS program, where new designs and new cost data are frequently offered for consideration. MONITOR can be run as an independent program, or it can be interfaced with the Waste System Transportation and Economic Simulation (WASTES) model, a program that simulates the movement of waste through a complete nuclear waste disposal system. The WASTES model drives the MONITOR model by providing it with the annual quantities of waste that are received, stored, and shipped at the MRS facility. Three runs of MONITOR are documented in this report. Two of the runs are for Version 1 of the MONITOR code. A simulation which uses the costs developed by the Ralph M. Parsons Company in the 2A (backup) version of the MRS cost estimate. In one of these runs MONITOR was run as an independent model, and in the other run MONITOR was run using an input file generated by the WASTES model. The two runs correspond to identical cases, and the fact that they gave identical results verified that the code performed the same calculations in both modes of operation. The third run was made for Version 2 of the MONITOR code. A simulation which uses the costs developed by the Ralph M. Parsons Company in the 2B (integral) version of the MRS cost estimate. This run was made with MONITOR being run as an ...
Date: December 1, 1986
Creator: Reimus, P. W.; Sevigny, N. L.; Schutz, M. E. & Heller, R. A.
Partner: UNT Libraries Government Documents Department

Chemical characterization, leach, and adsorption studies of solidified low-level wastes

Description: Laboratory and field leaching experiments are beig conducted by Pacific Northwest Laboratory (PNL) to investigate the performance of solidified low-level nuclear waste in a typical, arid, near-surface disposal site. Under PNL's Special Waste Form Lysimeters-Arid Program, a field test facility was constructed to monitor the leaching of commercial solidified waste. Laboratory experiments were conducted to investigate the leaching and adsorption characteristics of the waste forms in contact with soil. Liquid radioactive wastes solidified in cement, vinyl ester-styrene, and bitumen were obtained from commercial boiling water and pressurized water reactors, and buried in a field leaching facility on the Hanford site in southeastern Washington State. Batch leaching, soil column adsorption, and soil/waste form column experiments were conducted in the laboratory, using small-scale cement waste forms and Hanford site ground water. The purpose of these experiments is to evaluate the ability of laboratory leaching tests to predict leaching under actual field conditions and to determine which mechanisms (i.e., diffusion, solubility, adsorption) actually control the concentration of radionuclides in the soil surrounding the waste form. Chemical and radionuclide analyses performed on samples collected from the field and laboratory experiments indicate strong adsorption of /sup 134,137/Cs and /sup 85/Sr onto the Hanford site sediment. Small amounts of /sup 60/Co are leached from the waste forms as very mobile species. Some /sup 60/Co migrated through the soil at the same rate as water. Chemical constituents present in the reactor waste streams also found at elevated levels in the field and laboratory leachates include sodium, sulfate, magnesium, and nitrate. Plausible solid phases that could be controlling some of the chemical and radionuclide concentrations in the leachate were identified using the MINTEQ geochemical computer code.
Date: December 1, 1986
Creator: Walter, M.B.; Serne, R.J.; Jones, T.L. & McLaurine, S.B.
Partner: UNT Libraries Government Documents Department

Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

Description: This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions.
Date: December 1, 1986
Creator: Rector, D. R.; McCann, R. A.; Jenquin, U. P.; Heeb, C. M.; Creer, J. M. & Wheeler, C. L.
Partner: UNT Libraries Government Documents Department

[Circulation and exchange processes on the South Atlantic Bight Continental Shelf]. [Progress summary for 1986]

Description: A continuation of the physical oceanography program to investigate circulation and exchange processes on the South Atlantic Bight (SAB) Continental Shelf is proposed. The transport and dispersal of materials entering the inner shelf zone with river discharge is not well understood at present. Climatological data, satellite imagery, and numerical modeling results indicate two removal routes for these nearshore waters: northeast transport and offshore exchange between Cape Fear and Savannah during the spring and summer when maximum run-off and northward winds prevail; and southward transport and offshore exchange near Cape Canaveral during the fall when southward winds prevail. We have conducted interdisciplinary experiments to investigate the transport processes in the inner to outer shelf between Savannah, Georgia and Cape Fear, North Carolina. In addition we propose to continue synthesis and interpretation of current measurements. The analyses will focus on determining the coupling mechanisms of inner shelf and outer shelf waters with special emphasis placed on resolving the modes and rates of shelf water removal.
Date: December 31, 1986
Partner: UNT Libraries Government Documents Department

COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 3, Validation assessments

Description: This report presents the results of the COBRA-SFS (Spent Fuel Storage) computer code validation effort. COBRA-SFS, while refined and specialized for spent fuel storage system analyses, is a lumped-volume thermal-hydraulic analysis computer code that predicts temperature and velocity distributions in a wide variety of systems. Through comparisons of code predictions with spent fuel storage system test data, the code's mathematical, physical, and mechanistic models are assessed, and empirical relations defined. The six test cases used to validate the code and code models include single-assembly and multiassembly storage systems under a variety of fill media and system orientations and include unconsolidated and consolidated spent fuel. In its entirety, the test matrix investigates the contributions of convection, conduction, and radiation heat transfer in spent fuel storage systems. To demonstrate the code's performance for a wide variety of storage systems and conditions, comparisons of code predictions with data are made for 14 runs from the experimental data base. The cases selected exercise the important code models and code logic pathways and are representative of the types of simulations required for spent fuel storage system design and licensing safety analyses. For each test, a test description, a summary of the COBRA-SFS computational model, assumptions, and correlations employed are presented. For the cases selected, axial and radial temperature profile comparisons of code predictions with test data are provided, and conclusions drawn concerning the code models and the ability to predict the data and data trends. Comparisons of code predictions with test data demonstrate the ability of COBRA-SFS to successfully predict temperature distributions in unconsolidated or consolidated single and multiassembly spent fuel storage systems.
Date: December 1, 1986
Creator: Lombardo, N.J.; Cuta, J.M.; Michener, T.E.; Rector, D.R. & Wheeler, C.L.
Partner: UNT Libraries Government Documents Department

Evaluation of dense-phase ultrafine coal (DUC) as a fuel alternative for oil- and gas-designed boilers and heaters. Final report

Description: Utility and industrial firms currently using oil- and gas-fired boilers have an interest in substitution of coal for oil and gas as the primary boiler fuel. This interest stems from coal`s two main advantages over oil and gas-lower cost and security of supply. Recent efforts in the area of coal conversion have been directed to converting oil- and gas- fired boilers which were originally designed for coal-firing or were designed with some coal-firing capability. Boilers designed exclusively for oil- or gas-firing have not been considered viable candidates for coal conversion because they generally require a significant capacity derating and extensive and costly modifications. As a result, conversion of boilers in this class to coal-firing has generally been considered unattractive. Renewed interest in the prospects for converting boilers designed exclusively for oil- and gas-firing to coal firing has centered around the concept of using ``ultra fine`` coal as opposed to ``conventional grind`` pulverized coal. The main distinction being the finer particle size to which the former is ground. This fuel type may have characteristics which ameliorate many of the boiler problems normally associated with pulverized coal-firing. The overall concept for ultrafine coal utilization is based on a regional large preparation plant with distribution of a ready to fire fuel directly to many small users. This differs from normal practice in which final coal sizing is performed in pulverizers at the user`s site.
Date: December 1, 1986
Partner: UNT Libraries Government Documents Department