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Applications of energy-release-rate techniques to part-through cracks in experimental pressure vessels

Description: In nonlinear applications of computational fracture mechanics, energy release rate techniques are used increasingly for computing stress intensity parameters of crack configurations. Recently, deLorenzi used the virtual-crack-extension method to derive an analytical expression for the energy release rate that is better suited for three-dimensional calculations than the well-known J-integral. Certain studies of fracture phenomena, such as pressurized-thermal-shock of cracked structures, require that crack tip parameters be determined for combined thermal and mechanical loads. A method is proposed here that modifies the isothermal formulation of deLorenzi to account for thermal strains in cracked bodies. This combined thermo-mechanical formulation of the energy release rate is valid for general fracture, including nonplanar fracture, and applies to thermo-elastic as well as deformation plasticity material models. Two applications of the technique are described here. In the first, semi-elliptical surface cracks in an experimental test vessel are analyzed under elastic-plastic conditions using the finite element method. The second application is a thick-walled test vessel subjected to combined pressure and thermal shock loadings.
Date: January 1, 1982
Creator: Bass, B.R.; Bryan, R.H.; Bryson, J.W. & Merkle, J.G.
Partner: UNT Libraries Government Documents Department

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, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents 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. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.
Date: January 1, 1982
Creator: Cheverton, R.D.; Iskander, S.K. & Whitman, G.D.
Partner: UNT Libraries Government Documents Department

Compare containment subcompartment analysis code evaluation. [PWR; BWR]

Description: Nuclear power plant subcompartment analyses are required to determine the containment pressure distribution that might result from a loss-of-coolant accident. The pressure distribution is used to calculate structural and mechanical design loads. The COMPARE code is used widely to perform subcompartment analysis. However, several simplifying assumptions are utilized to facilitate solution of the complex transient, two-phase, multidimensional flow problem. In particular, it is assumed that the flow is homogeneous, in thermodynamic equilibrium, and one-dimensional. In this study, these assumptions are evaluated by performing simplified transport and relaxation analyses. This results in definition of (a) geometric features and early-time periods that produce significant deviations from reality and (b) specific areas that require further study.
Date: January 1, 1982
Creator: Gido, R.G. & Koestel, A.
Partner: UNT Libraries Government Documents Department

Calculational advance in the modeling of fuel-coolant interactions. [LMFBR]

Description: A new technique is applied to numerically simulate a fuel-coolant interaction. The technique is based on the ability to calculate separate space- and time-dependent velocities for each of the participating components. In the limiting case of a vapor explosion, this framework allows calculation of the pre-mixing phase of film boiling and interpenetration of the working fluid by hot liquid, which is required for extrapolating from experiments to a reactor hypothetical accident. Qualitative results are compared favorably to published experimental data where an iron-alumina mixture was poured into water. Differing results are predicted with LMFBR materials.
Date: January 1, 1982
Creator: Bohl, W.R.
Partner: UNT Libraries Government Documents Department

Steam-generator-tube-rupture transients for pressurized-water reactors

Description: Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures.
Date: January 1, 1982
Creator: Dobranich, D.; Henninger, R.J. & DeMuth, N.S.
Partner: UNT Libraries Government Documents Department

Fracture-mechanics data deduced from thermal-shock and related experiments with LWR pressure-vessel material

Description: Pressurized water reactors (PWRs) are susceptible to certain types of hypothetical accidents that can subject the reactor pressure vessel to severe thermal shock, that is, a rapid cooling of the inner surface of the vessel wall. The thermal-shock loading, coupled with the radiation-induced reduction in the material fracture toughness, introduces the possibility of propagation of preexistent flaws and what at one time were regarded as somewhat unique fracture-oriented conditions. Several postulated reactor accidents have been analyzed to discover flaw behavior trends; seven intermediate-scale thermal-shock experiments with steel cylinders have been conducted; and corresponding materials characterization studies have been performed. Flaw behavior trends and related fracture-mechanics data deduced from these studies are discussed.
Date: January 1, 1982
Creator: Cheverton, R.D.; Canonico, D.A.; Iskander, S.K.; Bolt, S.E.; Holz, P.P.; Nanstad, R.K. et al.
Partner: UNT Libraries Government Documents Department

SIMMER analysis of SRI postdisassembly expansion experiments. [LMFBR]

Description: Calculations of simulated postdisassembly expansions with SIMMER-II are compared to experimental data. The main features of the fluid motion are reproduced accurately. Good agreement is achieved for the important parameters of kinetic energy and impulse imparted to the vessel head. The most important calculated results are shown to be relatively insensitive to numerical modeling variations and uncertainties in many physical parameters. The geometric configuration of the upper internal structures is shown to affect significantly the vessel head impact energy and impulse.
Date: January 1, 1982
Creator: Bott, T.F. & Bell, C.R.
Partner: UNT Libraries Government Documents Department

Comparative evaluation of surface and downhole steam-generation techniques

Description: It has long been recognized that the application of heat to reservoirs containing high API gravity oils can substantially improve recovery. Although steam injection is currently the principal thermal recovery method, heat transmission losses associated with delivery of the steam from the surface generators to the oil-bearing formation has limited conventional steam injection to shallow reservoirs. The objective of the Department of Energy's Project DEEP STEAM is to develop the technology required to economically produce heavy oil from deep reservoirs. The tasks included in this effort are the development and evaluation of thermally efficient delivery systems and downhole steam generation systems. This paper compares the technical and economic performance of conventional surface steam drives, which are strongly influenced by heat losses, with (a) thermally efficient delivery (through insulated strings) of surface generated steam, (b) low pressure combustion downhole steam generation, (c) high pressure combustion downhole steam generation using air as the oxygen source, and (d) high pressure combustion downhole steam generation substituting pure oxygen for air. The selection of a preferred technology based upon either total efficiency or cost is found to be strongly influenced by reservoir depth, steam mass flow rate, and sandface steam quality. Therefore, a parametric analysis has been performed which examines varying depths, injection rates and steam qualities. Results indicate that the technologies are not readily distinguishable for low injectivity reservoirs in which conventional steam drives are feasible. However, high injection rates produce a notable cost difference between high pressure combustion systems and the other technologies. Issues that must be addressed before gaining further insight into the economic viability of downhole steam generation are discussed.
Date: January 1, 1982
Creator: Hart, C.
Partner: UNT Libraries Government Documents Department

Buckling of steel cylinders containing circular cutouts reinforced according to the area-replacement method

Description: The effect of the use of the area replacement method (ARM) for reinforcing circular penetrations in cylindrical steel shells has been studied both experimentally and analyticaly. How this type of reinforcement affects the buckling strength of a shell subjected to uniform axial compression is the specific area of investigation. In shells that are of such a quality that the penetration reduces the buckling strength, the use of the ARM will increase the bucking strength of the shell. In any case, the conservative knockdown factors suggested for buckling design by the American Society of Mechanical Engineer's (ASME) Boiler and Pressure Vessel Code should ensure an adequate margin to failure under this loading condition.
Date: January 1, 1982
Creator: Dove, R.C.; Bennett, J.G. & Butler, T.A.
Partner: UNT Libraries Government Documents Department

PRESS: a computer program for evaluating explosive-material loading processes

Description: The loading process for an explosive device must control within fixed tolerances not only the amount but the length of the explosive material that is compacted in the device's chargeholder. Most loading processes are empirically determined to achieve the desired tolerances with respect to the amount and length of the explosive material that is compacted in the chargeholder of the explosive device. An acceptable loading process for an explosive device can be established analytically by using the computer program, PRESS, which utilizes the known compaction behavior of the explosive material and a simple one-dimensional analysis to determine the effect of chargeholder friction and the applied compaction pressure on the final dimensions and the stress-density states of the compacted explosive material. PRESS is configured to evaluate the effects of unloading in addition to the effects of changes in the chargeholder dimensions, the applied compaction pressure and the coefficient of friction between the chargeholder and the explosive material. The details of the analysis that is incorporated into the computer program, PRESS, and an illustrative example of the results of a loading process parameter study are presented.
Date: January 1, 1982
Creator: Burchett, O.L.
Partner: UNT Libraries Government Documents Department

Prediction of failure modes for concrete nuclear-containment buildings. [PWR]

Description: The failure modes and associated failure pressures for two common generic types of PWR containments are predicted. One building type is a lightly reinforced, posttensioned structure represented by the Zion nuclear reactor containment. The other is the normally reinforced Indian Point containment. Two-dimensional models of the buildings developed using the finite element method are used to predict the failure modes and failure pressures. Predicted failure modes for both containments involve loss of structural integrity at the intersection of the cylindrical sidewall with the base slab.
Date: January 1, 1982
Creator: Butler, T.A.
Partner: UNT Libraries Government Documents Department

Comparisons of TRAC-PF-1 calculations with semiscale Mod-3 small-break tests S-SB-P1 and S-SB-P7. [PWR]

Description: Semiscale Tests S-SB-P1 and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Tests S-SB-P1 and S-SB-P7 simulated an equivalent pressurized-water-reactor (PWR) 2.5% communicative cold-leg break for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics.
Date: January 1, 1982
Creator: Sahota, M.S.
Partner: UNT Libraries Government Documents Department

Recovery of gas from hydrate deposits using conventional production technology. [Salt-frac technique]

Description: Methane hydrate gas could be a sizeable energy resource if methods can be devised to produce this gas economically. This paper examines two methods of producing gas from hydrate deposits by the injection of hot water or steam, and also examines the feasibility of hydraulic fracturing and pressure reduction as a hydrate gas production technique. A hydraulic fracturing technique suitable for hydrate reservoirs is also described.
Date: January 1, 1982
Creator: McGuire, P.L.
Partner: UNT Libraries Government Documents Department

Performance of the thermodynamic properties models in ASPEN. [Freon 12 and Freon 22]

Description: In the course of performing a number of analyses using ASPEN, the performance of the ASPEN models for computing thermodynamic properties has been observed. Pure-component properties for propane, isobutane, Freon 12 and Freon 22 and mixture properties for the propane-isobutane and the ethanol-water systems have been computed and the results compared with available data sources and with independent sources of computed properties. The built-in data regression system (DRS) of ASPEN was used to regress P-V-T and enthalpy departure data for isobutane to determine model-specific parameters. The extended Antoine vapor pressure parameters were calculated for Freon 12. The ethanol-water vapor-liquid equilibrium region was studied throughout the composition range for three isobaric data sets. Several activity coefficient models in ASPEN were fit to the data using various user-specified property routes.
Date: January 1, 1982
Creator: Fish, L.W. & Evans, D.R.
Partner: UNT Libraries Government Documents Department

Raft River 5MW Power Plant: a small binary power plant

Description: The Raft River 5MW power plant is a binary cycle pilot plant. The system uses isobutane in a dual boiling cycle. This cycle was selected because the well field and temperatures were not well known at the time of cycle selection, and therefore, a boiling cycle was desirable. The dual boiling feature provides about 15 to 20% more power and makes the output less sensitive to changes in geothermal temperature changes than a single boiler system. The plant design was based upon a 290F geothermal fluid temperature at the inlet to the plant and has a gross nominal generator rating of 5MW; however, actual output will vary according to ambient wet bulb temperatures over a range from 4.4MW to 6.2MW with the actual plant inlet temperature of 278F being obtained. The plant is supplied by three production wells. Geothermal fluid boost pumps within the plant inlet provide the pressure necessary to overcome plant pressure drop and return the fluid to the two injection sites. All long runs of the buried geothermal piping external to the plant boundaries use cement-asbestos pipe. The physical size and manpower requirements for the Raft River facility, the economics of small plant operation, and operational experience are discussed.
Date: January 1, 1982
Creator: Whitbeck, J.F.; DiBello, E.G. & Walrath, L.F.
Partner: UNT Libraries Government Documents Department

Neutron-exposure parameters for the fourth HSST series of metallurgical irradiation capsules

Description: The neutron exposure parameters for the Heavy Section Steel Technology (HSST) Experiments performed at the Oak Ridge National Laboratory (ORNL) can be determined conservatively to +-10% (1sigma) variance. The neutron exposure parameters used for this study were fluence greater than 1 MeV, fluence greater than 0.1 MeV, and displacements per atom (dpa). Measured reaction rates, calculated neutron transport fluxes, and cross sections values were combined in the logarithmic least square adjustment code LSL.
Date: January 1, 1982
Creator: Kam, F.B.K.; Stallmann, F.W.; Baldwin, C.A. & Fabry, A.
Partner: UNT Libraries Government Documents Department

Small rocket tornado probe

Description: A (less than 1 lb.) paper rock tornado probe was developed and deployed in an attempt to measure the pressure, temperature, ionization, and electric field variations along a trajectory penetrating a tornado funnel. The requirements of weight and materials were set by federal regulations and a one-meter resolution at a penetration velocity of close to Mach 1 was desired. These requirements were achieved by telemetering a strain gage transducer for pressure, micro size thermister and electric field, and ionization sensors via a pulse time telemetry to a receiver on board an aircraft that digitizes a signal and presents it to a Z80 microcomputer for recording on mini-floppy disk. Recording rate was 2 ms for 8 channels of information that also includes telemetry rf field strength, magnetic field for orientation on the rocket, zero reference voltage for the sensor op amps as well as the previously mentioned items also. The absolute pressure was recorded. Tactically, over 120 h were flown in a Cessna 210 in April and May 1981, and one tornado was encountered. Four rockets were fired at this tornado, missed, and there were many equipment problems. The equipment needs to be hardened and engineered to a significant degree, but it is believed that the feasibility of the probe, tactics, and launch platform for future tornado work has been proven. The logistics of thunderstorm chasing from a remote base in New Mexico is a major difficulty and reliability of the equipment another. Over 50 dummy rockets have been fired to prove trajectories, stability, and photographic capability. Over 25 electronically equipped rockets have been fired to prove sensors transmission, breakaway connections, etc. The pressure recovery factor was calibrated in the Air Force Academy blow-down tunnel. There is a need for more refined engineering and more logistic support.
Date: January 1, 1982
Creator: Colgate, S.A.
Partner: UNT Libraries Government Documents Department

Sentinel gap basalt reacted in a temperature gradient

Description: Six Basalt prisms were reacted in a controlled temperature gradient hydrothermal circulation system for two months. The prisms are centered at 72, 119, 161, 209, 270, and 310/sup 0/C. Total pressure is 1/3 kbar. All prisms show large weight loss: 5.5% to 14.9%. The matrix micropegmatite and natural nontronitic alteration readily react to clays at all temperatures. The first four prisms are coated with a Ca-smectite while the last two prisms are covered with discrete patches of K rich phengite and alkali feldspar. The clays may act as adsorbers of various ions.
Date: January 1, 1982
Creator: Charles, R.W. & Bayhurst, G.K.
Partner: UNT Libraries Government Documents Department

Post-test analysis of semiscale large-break test S-06-3 using TRAC-PF1. [PWR]

Description: The Transient Reactor Analysis Code (TRAC) is an advanced systems code for light-water-reactor accident analysis. The code was developed originally to analyze large-break loss-of-coolant accidents (LOCAs) and running time was not a primary development criterion. TRAC-PF1 was developed because increased application of the code to long transients such as small-break LOCAs required a faster-running code version. Although developed for long transients, its performance on large-break transients is still important. This paper assesses the ability of TRAC-PF1 to predict large-break-LOCA Test S-06-3 conducted in the Semiscale Mod-1 facility.
Date: January 1, 1982
Creator: Boyack, B.E.
Partner: UNT Libraries Government Documents Department

Verification of the three-dimensional thermal-hydraulic models of the TRAC accident-analysis code. [PWR]

Description: The Transient Reactor Analysis Code (TRAC) being developed at Los Alamos National Laboratory provides a best-estimate prediction of the response of light water reactors or test facilities to postulated accident sequences. One of the features of the code is the ability to analyze the vessel and its heated core in three dimensions. The code is being used to analyze the results of tests in a large-scale reflood test facility built in Japan, known as the Cylindrical Core Test Facility (CCTF). Two test runs have been analyzed that are useful for verification of the three-dimensional analysis capability of the TRAC code. One test began with an initial temperature skew across the heated core. The second test had a large radial power skew between the central and peripheral assemblies. The good agreement between the calculation and the experiment for both of these experiments demonstrates the three-dimensional analysis capability of the TRAC code.
Date: January 1, 1982
Creator: Motley, F.
Partner: UNT Libraries Government Documents Department

Analytical investigation of the applicability of simplified ratchetting and creep-fatigue rules to a nozzle-to-sphere geometry

Description: This paper presents an analysis of a nozzle-to-spherical-shell attachment and explores the applicability of simplified ratchetting and creep-fatigue rules to this attachment. A five-cycle inelastic analysis and creep-fatigue damage evaluation was carried out on this component. An elastic analysis also was done to provide input parameters required to apply the various rules and procedures of simplified analysis methods. Ten lines, or critical sections, were chosen for postprocessing to determine the ratchetting strain and creep-fatigue damage at both the inside and outside surfaces. At many of the 20 surface points analyzed, the inelastic analysis results did not develop a constant or decreasing pattern for the incremental strain or damage even after 5 cycles were analyzed. Failure to develop a constant or decreasing pattern was especially prevalent for creep damage. The results of the detailed inelastic analyses at the ten critical sections are compared with the results of elastic evaluations of ratchetting and creep-fatigue damage calculated according to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-47-13.
Date: January 1, 1982
Creator: Gwaltney, R.C.
Partner: UNT Libraries Government Documents Department

Calculations of the Westinghouse perturbation experiment at the Poolside Facility

Description: Discrete ordinate calculations are made and the results compared with measurements performed in the Poolside Facility for the purpose of validating various procedures adopted for the analysis of this facility. In addition, these calculations can be specifically used to verify the interpretation of measurements made to infer the perturbation effect of a Westinghouse surveillance capsule in a typical radiation environment. Comparisons indicate agreement on an absolute scale between measured and calculated reaction rates to within about 10% and agreement of the perturbation effect to within about 2%.
Date: January 1, 1982
Creator: Maerker, R.E. & Williams, M.L.
Partner: UNT Libraries Government Documents Department

Brief account of the effect of overcooling accidents on the integrity of PWR pressure vessels

Description: The occurrence in recent years of several (PWR) accident initiating events that could lead to severe thermal shock to the reactor pressure vessel, and the growing awareness that copper and nickel in the vessel material significantly enhance radiation damage in the vessel, have resulted in a reevaluation of pressure-vessel integrity during postulated overcooling accidents. Analyses indicate that the accidents of concern are those involving both thermal shock and pressure loadings, and that an accident similar to that at Rancho Seco in 1978 could, under some circumstances and at a time late in the normal life of the vessel, result in propagation of preexistent flaws in the vessel wall to the extent that they might completely penetrate the wall. More severe accidents have been postulated that would result in even shorter permissible lifetimes. However, the state-of-the-art fracture-mechanics analysis may contain excessive conservatism, and this possibility is being investigated. Furthermore, there are several remedial measures, such as fuel shuffling, to reduce the damage rate, and vessel annealing, to restore favorable material properties, that may be practical and used if necessary. 5 figures.
Date: January 1, 1982
Creator: Cheverton, R.D.
Partner: UNT Libraries Government Documents Department

Changing MFTF vacuum environment

Description: The Mirror Fusion Test Facility (MFTF) vaccum vessel will be about 60m long and 10m in diameter at the widest point. The allowable operating densities range from 2 x 10/sup 9/ to 5 x 10/sup 10/ particles per cc. The maximum leak rate of 10/sup -6/ tl/sec is dominated during operation by the deliberately injected cold gas of 250 tl/sec. This gas is pumped by over 1000 square meters of cryopanels, external sorbtion pumps and getters. The design and requirements have changed radically over the past several years, and they are still not in final form. The vacuum system design has also changed, but more slowly and less radically. This paper discusses the engineering effort necessary to meet these stringent and changing requirements. Much of the analysis of the internal systems has been carried out using a 3-D Monte Carlo computer code, which can estimate time dependent operational pressures. This code and its use will also be described.
Date: August 19, 1982
Creator: Margolies, D. & Valby, L.
Partner: UNT Libraries Government Documents Department