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Session 13: Technology Transfer of Geopressured/Geothermal Resources to Industry

Description: This research, conducted by the Bureau of Economic Geology and the Center for Energy Studies, includes the following areas of interest; geological studies depicting pressure gradients and thermal gradients, sand distribution and fault patterns, all of which are used in petroleum exploration; geophysical data for interpretation of seismic velocities based upon lithologic changes and subsurface discontinuities; sandstone consolidation data involving changes of permeabilities with depth and diagenetic histories of Cenozoic rocks in the Gulf Coast Basin--this work also covers fluid migration pathways and resulting rock-water interactions and has led to a better understanding of generation, maturation and accumulation of hydrocarbons; work on salinity of formation waters covering several areas of study, such as chemical analysis to anticipate scale and corrosion problems, and investigations of logging techniques to better ascertain salinity of use of well logs; reservoir continuity studies, together with computational modeling to assist in estimation of ultimate recoveries and formation drives; rock mechanics studies, which have recently led to the development of new models to account for creep and determine compressibilities of sandstones and shales in geopressured environments; co-production of gas and water in watered-out gas reservoirs.
Date: December 1, 1983
Creator: Dorfman, Myron H. & Morton, Robert A.
Partner: UNT Libraries Government Documents Department

Factors affecting the integrity of PWR pressure vessels during overcooling accidents

Description: The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, if certain postulated accidents, referred to as overcooling accidents, were to occur, the pressure vessel could be subjected to severe thermal shock while the pressure is substantial. As a result, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner-surface flaws prior to the vessel's normal end of life. A fracture-mechanics analysis for a typical postulated accident and also related thermal-shock experiments indicate that very shallow surface flaws that extend through the cladding into the base material could propagate. This is of particular concern because shallow flaws appear to be the most probable and presumably are the most difficult to detect.
Date: January 1, 1983
Creator: Cheverton, R.D.
Partner: UNT Libraries Government Documents Department

X-ray line broadening studies on aluminum nitride, titanium carbide and titanium diboride modified by high pressure shock loading

Description: Powders of AlN, TiC and TiB/sub 2/ have been subjected to controlled shock loading with peak pressures in the samples between 14 to 27 GPa and preserved for post-shock study. Broadened x-ray diffraction peak profiles are analyzed by a simplified method and show increases in residual lattice strain and small decreases in crystallite size. Strain values range from 10/sup -5/ to 10/sup -4/ for TiB/sub 2/ and to values larger than 10/sup -3/ for TiC and AlN.
Date: January 1, 1983
Creator: Morosin, B. & Graham, R.A.
Partner: UNT Libraries Government Documents Department

Prestressed-concrete pressure vessels and their applicability to advanced-energy-system concepts

Description: Prestressed concrete pressure vessels (PCPVs) are, in essence, spaced steel structures since their strength is derived from a multitude of steel elements made up of deformed reinforcing bars and prestressing tendons which are present in sufficient quantities to carry tension loads imposed on the vessel. Other major components of a PCPV include the concrete, liner and cooling system, and insulation. PCPVs exhibit a number of advantages which make them ideally suited for application to advanced energy concepts: fabricability in virtually any size and shape using available technology, improved safety, reduced capital costs, and a history of proven performance. PCPVs have many applications to both nuclear- and non-nuclear-based energy systems concepts. Several of these concepts will be discussed as well as the research and development activities conducted at ORNL in support of PCPV development.
Date: January 1, 1983
Creator: Naus, D.J
Partner: UNT Libraries Government Documents Department

TRAC analysis and support of Oconee-1 PTS studies. [Pressurized thermal shock following overcooling transients]

Description: This paper describes the overall pressurized thermal shock (PTS) program at Los Alamos with emphasis on TRAC-PF1 calculations of severe overcooling transients for the Oconee-1 pressurized water reactor (PWR). A summary of results for several calculations are presented for the Oconee-1 PWR along with detailed discussions of two of the most severe overcooling transients predicted (main steam-line break and turbine-bypass valve (TBV) failures). The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to a detailed model of the plant Integrated Control System (ICS). For the Oconee-1 main steam-line break transient, a minimum downcomer fluid temperature of approx. 405 K was predicted. For the TBV transient involving the failure of one bank of TBVs to close after initially opening following reactor and turbine trips, an extrapolated downcomer fluid temperature of approx. 365 K was estimated. The latter temperature is at the nil-ductility temperature (NDT) limit (approx. 365 K) for Oconee-1.
Date: January 1, 1983
Creator: Ireland, J.R.
Partner: UNT Libraries Government Documents Department

TRAC calculations of overcooling transients in PWRs for pressurized thermal-shock analysis

Description: This paper briefly describes the overall pressurized thermal shock (PTS) program at Los Alamos with emphasis on TRAC-PF1 calculations of severe overcooling transients in pressurized water reactors (PWRs). Overcooling transients for both the Oconee-1 and Calvert Cliffs-1 nuclear plants have been performed. A summary of results for several calculations are presented for the Oconee-1 PWR along with detailed discussions of two of the most severe overcooling transients predicted (main steam-line break and turbine-bypass valve (TBV) failures). The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to a detailed model of the plant integrated control system (ICS). For the Oconee-1 main steam-line break transient, a minimum downcomer fluid temperature of approx. 405/sup 0/K was predicted. For the transient involving the failure of one bank of TBVs to close after initially opening following reactor and turbine trips, and extrapolated downcomer fluid temperature of approx. 365/sup 0/K was estimated. The latter temperature is at the nil-ductility temperature (NDT) limit (approx. 365/sup 0/K) for Oconee-1.
Date: January 1, 1983
Creator: Ireland, J.R. & Boyack, B.E.
Partner: UNT Libraries Government Documents Department

Pressure transient method for front tracking

Description: A pressure transient technique for tracking the advance of cold water fronts during water flooding and goethermal injection operations has been developed. The technique is based on the concept that the steady state pressure buildup in the reservoir region inside the front can be calculated by a fluid skin factor. By analyzing successive pressure falloff tests, the advance of the front in the reservoir can be monitored. The validity of the methods is demonstrated by application to three numerically simulated data sets, a nonisothermal step-rate injection test, a series of pressure falloffs in a multilayered reservoir, and a series of pressure falloff tests in a water flooded oil reservoir.
Date: August 1, 1983
Creator: Benson, S.M. & Bodvarsson, G.S.
Partner: UNT Libraries Government Documents Department

Generating capacity of the Heber geothermal field, California

Description: Using numerical simulation techniques and the radial model developed for the study of the natural state of the Heber field (Lippmann and Bodvarsson, 1983b), the response of this geothermal system to exploitation is analyzed. In this study the generation rate in the field is allowed to build up over a period of 10 years; after that, 30 years of constant power production is assumed. Full (100%) injection of the spent brines is considered, the fluids being injected 2250 m (near injection) or 4250 m (far injection) from the center of the system. The study shows that a maximum of 6000 kg/s (equivalent to approximately 300 MW/sub e/) of fluids may be produced for the near injection case, but only 3000 kg/s (equivalent to approximately 150 MW/sub e/) for the far injection case. The results indicate that the possible extraction rates (generating capacity) generally are limited by the pressure drop in the reservoir. The average temperature of the produced fluids will decline 10 to 18/sup 0/C over the 40-year period.
Date: December 1, 1983
Creator: Lippmann, M.J. & Bodvarsson, G.S.
Partner: UNT Libraries Government Documents Department

Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

Description: The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs.
Date: January 1, 1983
Creator: Kroeger, P.G.; Colman, J. & Araj, K.
Partner: UNT Libraries Government Documents Department

Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

Description: In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization.
Date: January 1, 1983
Creator: Kroeger, P.G.; Colman, J. & Hsu, C.J.
Partner: UNT Libraries Government Documents Department

Quality assurance of PTS thermal hydraulic calculations at BNL. [Pressurized Thermal Shock]

Description: Rapid cooling of the reactor pressure vessel at high pressure has a potential of challenging the vessel integrity. This phenomenon is called overcooling or Pressurized Thermal Shock (PTS). The Nuclear Regulatory Commission (NRC) has selected three plants representing three types of PWRs in use for detailed PTS study. Oconee-1 (B and W), Calvert Cliffs (C.E.), and H.B. Robinson (Westinghouse). The Brookhaven National Laboratory (BNL) has been requested by NRC to review and compare the input decks developed at LANL and INEL, and to compare and explain the differences between the common calculations performed at these two laboratories. However, for the transients that will be computed by only one laboratory, a consistency check will be performed. So far only Oconee-1 calculations have been reviewed at BNL, and the results are presented here.
Date: January 1, 1983
Creator: Rohatgi, U.S.; Pu, J.; Jo, J. & Saha, P.
Partner: UNT Libraries Government Documents Department

Design and construction of the vacuum vessel for the tandem Mirror Fusion Test Facility

Description: We have designed the MFTF-B vacuum vessel both to maintain the required vacuum environment and to structurally support the 42 superconducting magnets plus auxiliary internal and external equipment. The design calculations were greatly aided by computer models, which also speeded our redesign effort when the machine configuration was changed to the Axicelll MFTF-B this past year. Our field construction and erection effort should meet the July 1984 completion date for the vacuum vessel.
Date: August 12, 1983
Creator: Gerich, J.W.
Partner: UNT Libraries Government Documents Department

Dislocation kinetics and the acoustic-wave approximation for liquids

Description: A dislocation-dependent model for liquids describes the lattice deformation and the fluidity deformation as additive deformations. The lattice deformation represents distortions of an atom's potential energy structure and is a recoverable deformation response. The fluidity deformation represents discontinuous repositioning of atoms by dislocation kinetics in the lattice structure and is a nonrecoverable deformation response. From this model, one concludes that in liquids the acoustic-wave approximation is a description of a recoverable oscillation deformation that has dissipation because of dislocation kinetics. Other more-complex waves may exist, but such waves would rapidly disappear because of the small thermodynamic potential for dislocation kinetics in liquids.
Date: March 1, 1983
Creator: Stout, R.B.
Partner: UNT Libraries Government Documents Department

Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem-mirror fusion reactor

Description: A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li/sub 2/O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail.
Date: January 10, 1983
Creator: Werner, R.W.; Hoffman, M.A. & Johnson, G.L.
Partner: UNT Libraries Government Documents Department

Nonlinear fluid/structure interaction relating a rupture-disc pressure-relief device. [LMFBR]

Description: Rupture disc assemblies are used in piping network systems as a pressure-relief device. The reverse-buckling type is chosen for application in a liquid metal fast breeder reactor. This assembly is used successfully in systems in which the fluid is highly compressible, such as air; the opening up of the disc by the knife setup is complete. However, this is not true for a liquid system; it had been observed experimentally that the disc may open up only partially or not at all. Therefore, to realistically understand and represent a rupture disc assembly in a liquid environment, the fluid-structure interactions between the liquid medium and the disc assembly must be considered. The methods for analyzing the fluid and the disc and the mechanism interconnecting them are presented. The fluid is allowed to cavitate through a column-cavitation model and the disc is allowed to become plastically deformed through the classic Von Mises' yield criteria, when necessary.
Date: January 1, 1983
Creator: Hsieh, B.J.; Kot, C.A.; Shin, Y.W. & Youngdahl, C.K.
Partner: UNT Libraries Government Documents Department

Calculation of the limiting CESSAR Feedwater Line-Break and Steam Line-Break transients

Description: Argonne National Laboratory (ANL), under contract to the Nuclear Regulatory Commission, performed audit calculations of the limiting Feedwater Line Break (FLB) and Steam Line Break (SLB) transients presented in the CESSAR FSAR. The results of the FLB and SLB calculations are discussed.
Date: January 1, 1983
Creator: Peeler, G.B.; Kennedy, M.F.; Caraher, D.L.; Guttmann, J. & Chung, K.S.
Partner: UNT Libraries Government Documents Department

Role of pressure in understanding the anomalous superconductivity in europium (molybdenum)/sub 6/(sulfur)/sub 8/ and (TMTSF)/sub 2/FSO/sub 3/

Description: Both the Chevrel phase compound EuMo/sub 6/S/sub 8/ and the organic material, (TMTSF)/sub 2/FSO/sub 3/ are superconducting only under moderate pressure. In both instances the absence of superconductivity at ambient pressure is directly attributed to a low temperature structural distortion that introduces a gap over all or part of the Fermi surface. The role of pressure is to suppress the transition and thus allow the electrons to condense into the superconducting state. In EuMo/sub 6/S/sub 8/, details of the pressure dependence of both the structural and superconducting transition have been explained on the basis of a competition between a charge density wave-type state and superconductivity. In the case of (TMTSF)/sub 2/FSO/sub 3/ an anion ordering giving rise to a metal-insulator transition is responsible for suppressing superconductivity. The critical magnetic fields of EuMo/sub 6/S/sub 8/ are extremely anomalous and are related to the magnetism of the Eu as well as the structure of the compound.
Date: January 1, 1983
Creator: Wolf, S.A.; Huang, C.Y.; Lacoe, R.C.; Chaikin, P.M.; Fuller, W.W.; Luo, H.L. et al.
Partner: UNT Libraries Government Documents Department

Mechanical interactions of UIS support columns. [LMFBR]

Description: Code development involving above-core structures (ACS) has recently focused on modeling the complexities of mechanical interactions in the ACS support columns which play a very important role in their behavior. These developments are directed toward two considerations: (1) the prediction of the forces exerted by the column in a hypothetical core-disruptive accident (HCDA) in order that the motion of the ACS can be predicted in a coupled fluid-structure analysis, (2) the calculation of the strains and deformations of the support columns so that situations which lead to complete failure can be identified. Finite element capabilities have been developed to handle various types of plant design for the analysis of coupled hydrodynamics and structural response. Beam elements, which previously represented the support columns were able to account for geometric nonlinearities and material nonlinearities, however, changes in the column cross section were not treated. Therefore, one of the aims of this study was to examine the effect of the change in cross section on the behavior of the support columns. A second effect which has been studied is the behavior of support columns consisting of two concentric cylinders.
Date: January 1, 1983
Creator: Kennedy, J.M. & Belytschko, T.B.
Partner: UNT Libraries Government Documents Department

Dynamic response of a lare loop-type LMFBR head closure to HCDA loads

Description: An investigation is presented here on the dynamic structural response of the primary vessel's head closure to slug impact loadings generated from a 1000 MJ source term. The reference reactor considered was designed in a loop configuration. The head structure consisted of a deck and a triple rotatable plug assembly. The deck was a large annular structure that was supported along its outer periphery by the vessel flange. The deck provided support to the triple rotatable plug (TRP) assembly along its inner periphery. Two designs were considered for the deck structure: a reference design and an alternate design. The reference deck was designed as a single flat annular plate. For the alternate design, the deck plate was reinforced by adding an extender cyclinder with a flange and flanged webs between the deck-plate and cylinder. A decoupled analysis of this fluid-structure interaction problem was performed. A two-dimensional axisymmetric hydrodynamic computation (reported in a companion paper) was performed to define the head loadings. Then a three-dimensional structural response computation was made to assess the containment capability of the head closure.
Date: January 1, 1983
Creator: Kulak, R.F. & Fiala, C.
Partner: UNT Libraries Government Documents Department

Dynamic analysis of large suspended LMFBR reactor vessels

Description: Large breeder reactor vessels are often designed under the top-suspended condition. Since the vessel contains a large volume of liquid sodium as reactor coolant, the structural integrity of the vessel bottom head and its effect on the vessel dynamic response are of great importance to the safety and reliability of the reactor systems. This paper presents a dynamic analysis of the large suspended reactor vessel subjected to the horizontal earthquake excitation with the emphasis on the effect of bottom head vibration on fluid pressure and sloshing response. Unlike the conventional lumped mass method, the present analysis treats the liquid sodium as a continuum medium. As a result, the important effects ignored in the lumped mass method such as fluid coupling, fluid-structure interaction, interaction between sloshing and vessel vibration, etc. can be accounted into the analysis.
Date: January 1, 1983
Creator: Ma, D.C.; Gvildys, J. & Chang, Y.W.
Partner: UNT Libraries Government Documents Department

Estimation of structural reliability under combined loads. [PWR; BWR]

Description: For the overall safety evaluation of seismic category I structures subjected to various load combinations, a quantitative measure of the structural reliability in terms of a limit state probability can be conveniently used. For this purpose, the reliability analysis method for dynamic loads, which has recently been developed by the authors, was combined with the existing standard reliability analysis procedure for static and quasi-static loads. The significant parameters that enter into the analysis are: the rate at which each load (dead load, accidental internal pressure, earthquake, etc.) will occur, its duration and intensity. All these parameters are basically random variables for most of the loads to be considered. For dynamic loads, the overall intensity is usually characterized not only by their dynamic components but also by their static components. The structure considered in the present paper is a reinforced concrete containment structure subjected to various static and dynamic loads such as dead loads, accidental pressure, earthquake acceleration, etc. Computations are performed to evaluate the limit state probabilities under each load combination separately and also under all possible combinations of such loads.
Date: January 1, 1983
Creator: Shinozuka, M.; Kako, T.; Hwang, H.; Brown, P. & Reich, M.
Partner: UNT Libraries Government Documents Department

Theory and application of a three-dimensional code SHAPS to complex piping systems. [LMFBR]

Description: This paper describes the theory and application of a three-dimensional computer code SHAPS to the complex piping systems. The code utilizes a two-dimensional implicit Eulerian method for the hydrodynamic analysis together with a three-dimensional elastic-plastic finite-element program for the structural calculation. A three-dimensional pipe element with eight degrees of freedom is employed to account for the hoop, flexural, axial, and the torsional mode of the piping system. In the SHAPS analysis the hydrodynamic equations are modified to include the global piping motion. Coupling between fluid and structure is achieved by enforcing the free-slip boundary conditions. Also, the response of the piping network generated by the seismic excitation can be included. A thermal transient capability is also provided in SHAPS. To illustrate the methodology, many sample problems dealing with the hydrodynamic, structural, and thermal analyses of reactor-piping systems are given. Validation of the SHAPS code with experimental data is also presented.
Date: January 1, 1983
Creator: Wang, C.Y.
Partner: UNT Libraries Government Documents Department

HCDA behavior within the primary containment of a large pool-type LMFBR

Description: This paper discusses, specifically, the primary containment response of a large pool-type reactor under HCDA conditions. A large-diameter, thin-walled, pool-type reactor may have some inherent safety advantages in terms of energy accommodation during HCDA loads. The purpose of this study was to predict the containment response from the energetic excursion and to determine the impact loading on the reactor deck for a more detailed analysis. The essential features of the primary system which were modelled include the reactor core, radial shield, redan (separating the hot and cold pools), core support structure (CSS), upper internal structure (UIS), the sodium coolant, and the reactor vessel. Three different UIS configurations were studied and the effects of flow paths on the primary containment response were noted. In each case the primary systems were identical except for the upper internal structure (UIS) which was parametrically varied to simulate; (1) annular flow, (2) a horizontal guide plate, and (3) a cylindrical shroud with guide plate.
Date: January 1, 1983
Creator: Zeuch, W.R.; Wang, C.Y. & Seidensticker, R.W.
Partner: UNT Libraries Government Documents Department

Nonlinear finite-element analysis of a reinforced-concrete Mark III containment under pressure and gravity loads. [BWR]

Description: An analysis of a Mark III reactor containment vessel subjected to a uniformly increasing internal pressure and gravity loads is carried out in order to ascertain the load carrying capacity of the structure under hydrogen burn. The analysis is conducted by using a nonlinear finite element model that includes nonlinearities in the strain-displacement relations as well as in the material constitutive equations. In this analysis, the nonlinear behavior of the liner and reinforcement steels is described by a von Mises elastic-plastic model with isotropic hardening. A recently developed elastic-plastic-fracture model that includes both the cracking and crushing limit states is used for the plain concrete. Consistent smearing and de-smearing procedures are then used to represent the composite material properties of the reinforced concrete by an anisotropic and locally homogeneous continuum. Results pertaining to the critical regions of the containment where cracking of the concrete, yielding of the reinforcement bars, and substantial straining of the liner take place are discussed.
Date: January 1, 1983
Creator: Sharma, S.; Reich, M.; Shteyngart, S. & Chang, T.Y.
Partner: UNT Libraries Government Documents Department