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Graphite design handbook

Description: The objectives of the Graphite Design Handbook (GDH) are to provide and maintain a single source of graphite properties and phenomenological model of mechanical behavior to be used for design of MHTGR graphite components of the Reactor System, namely, core support, permanent side reflector, hexagonal reflector elements, and prismatic fuel elements; to provide a single source of data and material models for use in MHTGR graphite component design, performance, and safety analyses; to present properties and equations representing material models in a form which can be directly used by the designer or analyst without the need for interpretation and is compatible with analytical methods and structural criteria used in the MHTGR project, and to control the properties and material models used in the MHTGR design and analysis to proper Quality Assurance standards and project requirements. The reference graphite in the reactor internal components is the nuclear grade 2020. There are two subgrades of interest, the cylinder nuclear grade and the large rectangular nuclear grade. The large rectangular nuclear grade is molded in large rectangular blocks. It is the reference material for the permanent side reflector and the central column support structure. The cylindrical nuclear grade is isostatically pressed and is intended for use as the core support component. This report gives the design properties for both H-451 and 2020 graphite as they apply to their respective criteria. The properties are presented in a form for design, performance, and safety calculations that define or validate the component design. 103 refs., 20 figs., 19 tabs.
Date: September 1, 1988
Creator: Ho, F.H.
Partner: UNT Libraries Government Documents Department

Design requirements for high-temperature metallic component materials in the US modular HTGR

Description: The modular high temperature gas-cooled reactor (MHTGR) is a 350 MW(t) second generation reactor system design which during normal operation circulates helium with a mixed mean coal and hot temperature of 260/sup 0/C (500/sup 0/C) and 690/sup 0/C (1270/sup 0/F), respectively. The design incorporates passive design features which allow the plant to be safely shutdown and cooled with no active systems or operator action being required. A key feature of this concept is the capability of the residual heat removal by passive conduction cooldown from the core to the reactor cavity via an uninsulated vessel. The MHTGR uses a number of metallic components. A description of these components and their design requirements are presented in this paper.
Date: June 1, 1988
Creator: Shenoy, A.S. & Betts, W.S.
Partner: UNT Libraries Government Documents Department

Analytical Chemistry Laboratory: Progress report for FY 1988

Description: The purpose of this report is to summarize the activities of the Analytical Chemistry Laboratory (ACL) at Argonne National Laboratory (ANL) for fiscal year 1988 (October 1987 through September 1988). The Analytical Chemistry Laboratory is a full-cost recovery service center, with the primary mission of providing a broad range of analytical chemistry support services to the scientific and engineering programs at ANL. In addition, the ACL conducts a research program in analytical chemistry, works on instrumental and methods development, and provides analytical services for governmental, educational, and industrial organizations. The ACL handles a wide range of analytical problems, from routine standard analyses to unique problems that require significant development of methods and techniques.
Date: December 1, 1988
Creator: Green, D.W.; Heinrich, R.R.; Graczyk, D.G.; Lindahl, P.C. & Erickson, M.D.
Partner: UNT Libraries Government Documents Department

Steam generator group project: Task 13 final report: Nondestructive examination validation

Description: The Steam Generator Group Project (SGGP) was a multi-task effort using the retired-from-service Surry 2A pressurized water reactor steam generator as a test bed to investigate the reliability and effectiveness of in-service nondestructive eddy current (EC) inspection equipment and procedures. The information developed provided the technical basis for recommendations for improved in- service inspection and tube plugging criteria of steam generators. This report describes the results and analysis from Task 13--NDE Validation. The primary objective of Task 13 was to validate the EC inspection to detect and size tube defects. Additional objectives were to assess the nature and severity of tube degradation from all regions of the generator and to measure the remaining integrity of degraded specimens by burst testing. More than 550 specimens were removed from the generator and included in the validation studies. The bases for selecting the specimens and the methods and procedures used for specimen removal from the generator are reported. Results from metallurgical examinations of these specimens are presented and discussed. These examinations include visual inspection of all specimens to locate and identify tube degradation, metallographic examination of selected specimens to establish defect severity and burst testing of selected specimens to establish the remaining integrity of service-degraded tubes. Statistical analysis of the combined metallurgical and EC data to determine the probability of detection (POD) and sizing accuracy are reported along with a discussion of the factors which influenced the EC results. Finally, listings of the metallurgical and corresponding EC data bases are given. 12 refs., 141 figs., 24 tabs.
Date: August 1, 1988
Creator: Bradley, E.R.; Doctor, P.G.; Ferris, R.H. & Buchanan, J.A.
Partner: UNT Libraries Government Documents Department

Reservoir and injection technology: Geothermal reservoir engineering research at Stanford: Third annual report for the period October 1, 1986 through September 30, 1987: (Final report)

Description: This paper discusses different aspects of geothermal reservoir engineering. General topics covered are: reinjection technology, reservoir technology, and heat extraction. (LSP)
Date: February 1, 1988
Creator: Ramey, H.J. Jr.; Horne, R.N.; Miller, F.G. & Brigham, W.E.
Partner: UNT Libraries Government Documents Department

Energy Program annual report

Description: The national economy is particularly dependent on efficient electrical generation and transportation. Electrical demand continues to grow and will increasingly rely on coal and nuclear fuels. The nuclear power industry still has not found a solution to the problem of disposing of the waste produced by nuclear reactors. Although coal is in ample supply and the infrastructure is in place for its utilization, environmental problems and improved conversion processes remain technical challenges. In the case of transportation, the nation depends almost exclusively on liquid fuels with attendant reliance on imported oil. Economic alternates---synfuels from coal, natural gas, and oil shale, or fuel cells and batteries---have yet to be developed or perfected so as to impact the marketplace. Inefficiencies in energy conversion in almost all phases of resource utilization remain. These collective problems are the focus of the Energy Program.
Date: February 1, 1988
Creator: Borg, I.Y. (ed.)
Partner: UNT Libraries Government Documents Department

A method for suppression of pressure pulses in fluid-filled piping: Theoretical analysis

Description: A simple, nondestructive method to suppress pressure pulses in a fluid-filled piping is theoretically analyzed, and the result provides the basis needed for design and evaluation of a pressure-pulse suppression device based on the proposed theory. The method is based on forming of fluid jets in the event of a pressure surge such that the pulse height as well as the energy of the pulse are reduced. The result for pressure pulses in the range of practical interest shows that a substantial reduction can be attained in the pulse height with accompanied reduction of pulse energy remaining in the system. The analysis also reveals that a certain amount of trade-off exists in the design of the suppression device; a certain level of pulse energy remaining in the system must be accepted in order to limit the pulse height below a certain level and vice versa. 7 refs., 5 figs.
Date: June 1, 1988
Creator: Shin, Y.W. & Wiedermann, A.H.
Partner: UNT Libraries Government Documents Department

Once-through steam-generator sensitivity calculations

Description: A series of TRAC-PF1/MOD2 thermal-hydraulic calculations has been performed to determine the effect of uncertainties in modeling once through steam-generator (OTSG) secondary-side phenomena on the calculated behavior of Babcock and Wilcox power plants. The calculations were performed by varying parameters in correlations for the secondary-side phenomena. The parameters and transients were chosen to show the maximum expected sensitivity of the calculated results to the parameter variations. The parameters were then varied over a range representing the estimated uncertainty in the correlation. In this manner, the sensitivity if the calculated plant behavior to the modeling uncertainties was determined with a reasonable number of calculations. The sensitivity of calculated plant behavior to variations in interfacial heat-transfer in the OTSG secondaries was determined in a series of steam-generator overfill transient calculations. Calculations were performed for a main steam line break (MSLB) transient to quantify the sensitivity to variations in interfacial drag in the secondaries; the interfacial drag was varied in these calculations to indicate the effects of entrainment and de-entrainment processes, for which no specific models exist in the code. In addition to the transient calculations, a series of steady-state calculations was performed to determine the sensitivity of the OTSG primary-to-secondary heat transfer to the assumed fraction of tubes wetted by the auxiliary feedwater (AFW) injection. The plant model used for the sensitivity calculations was qualified by performing a benchmark calculation for a natural circulation test in the TMI-1 plant.
Date: January 1, 1988
Creator: Steiner, J.L. & Siebe, D.A.
Partner: UNT Libraries Government Documents Department

TRAC PF1/MOD1 calculations and data comparisons for mist feed and bleed and steam generator tube rupture experiments

Description: Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents, loss of feedwater and other transients in Babcock and Wilcox (BandW) plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 /times/ 4 (two hot legs and steam generators (SGs), four cold legs and reactor coolant pumps) representation of lowered-loop reactor system of the BandW design. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Two other integral experimental facilities are included in the IST program: test loops at the University of Maryland, College Park, and at SRI International (SRI-2). The objective of the IST tests is to generate high-quality experimental data to be used for assessing thermal-hydraulic safety computer codes. Efforts are under way at Los Alamos to assess TRAC-PF1/MOD1 against data from each of the IST facilities. Calculations and data comparisons for TRAC-PF1/MOD1 assessment are presented for two transients run in the MIST facility. These are MIST Test 330302, a feed and bleed test with delayed high-pressure injection; and Test 3404AA, an SG tube-rupture test with the affected SG isolated. Only MIST assessment results are presented in this paper. The TRAC-PF1/MOD1 calculations completed to date for MIST tests are in reasonable agreement with the data from these tests. Reasonable agreement is defined as meaning that major trends are predicted correctly, although TRAC values are frequently outside the range of data uncertainty. We believe that correct conclusions will be reached if the code is used in similar applications despite minor code/model deficiencies. 7 refs., 5 figs., 2 tabs.
Date: January 1, 1988
Creator: Siebe, D.A.; Boyack, B.E. & Steiner, J.L.
Partner: UNT Libraries Government Documents Department

Reliability of the Solar One plant during the power production phase: August 1, 1984--July 31, 1987

Description: The power production phase at Solar One spanned three years from August 1, 1984 through July 31, 1987. In that period the plant achieved an average availability, during hours of sunshine, of 81.7%. This report presents the frequencies and causes of the plant outages that occurred. The eleven most important causes composed 75% of the total outage time. Qualitative insights related to the origin and mitigation of these causes are provided. Also presented are insights and statistics regarding the reliability of the heliostat field. The quantitative and qualitative information presented in this report will be useful to studies aimed at improving the reliability of future solar central receiver power plants. 20 refs., 11 figs., 3 tabs.
Date: October 1, 1988
Creator: Kolb, G.J. & Lopez, C.W.
Partner: UNT Libraries Government Documents Department

Interpretation of Self-Potential Anomalies Using Constitutive Relationships for Electrochemical and Thermoelectric Coupling Coefficients

Description: Constitutive relationships for electrochemical and thermoelectric cross-coupling coefficients are derived using ionic mobilities, applying a general derivative of chemical potential and employing the zero net current condition. The general derivative of chemical potential permits thermal variations which give rise to the thermoelectric effect. It also accounts for nonideal solution behavior. An equation describing electric field strength is similarly derived with the additional assumption of electrical neutrality in the fluid Planck approximation. The Planck approximation implies that self-potential (SP) is caused only by local sources and also that the electric field strength has only first order spatial variations. The derived relationships are applied to the NaCl-KCl concentration cell with predicted and measured voltages agreeing within 0.4 mV. The relationships are also applied to the Long Valley and Yellowstone geothermal systems. There is a high degree of correlation between predicted and measured SP response for both systems, giving supporting evidence for the validity of the approach. Predicted SP amplitude exceeds measured in both cases; this is a possible consequence of the Planck approximation. Electrochemical sources account for more than 90% of the predicted response in both cases while thermoelectric mechanisms account for the remaining 10%; electrokinetic effects are not considered. Predicted electrochemical and thermoelectric voltage coupling coefficients are comparable to values measured in the laboratory. The derived relationships are also applied to arbitrary distributions of temperature and fluid composition to investigate the geometric diversity of observed SP anomalies. Amplitudes predicted for hypothetical saline spring and hot spring environments are less than 40 mV. In contrast, hypothetical near surface steam zones generate very large amplitudes, over 2 V in one case. These results should be viewed with some caution due to the uncertain validity of the Planck approximation for these conditions. All amplitudes are controlled by electrochemical mechanisms. Polarities are controlled by the ...
Date: January 1, 1988
Creator: Knapp, R. B. & Kasameyer, P. W.
Partner: UNT Libraries Government Documents Department

Steam generator tube integrity program: Phase II, Final report

Description: The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted.
Date: August 1, 1988
Creator: Kurtz, R. J.; Bickford, R. L.; Clark, R. A.; Morris, C. J.; Simonen, F. A. & Wheeler, K. R.
Partner: UNT Libraries Government Documents Department

Technology, safety and costs of decommissioning a reference pressurized water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

Description: Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies on conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference pressurized water reactor (PWR) described in the earlier study; defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs; and completing a study of recent PWR steam generator replacements to determine realistic estimates for time, costs and doses associated with steam generator removal during decommissioning. This report presents the results of recent PNL studies to provide supporting information in four areas concerning decommissioning of the reference PWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; assessing the cost and dose impacts of recent steam generator replacements; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.
Date: July 1, 1988
Creator: Konzek, G. J. & Smith, R. I.
Partner: UNT Libraries Government Documents Department

Environmentally assisted cracking in light water reactors

Description: Research during the past year focused on (1) stress corrosion cracking (SCC) of austentitic stainless steels (SS), (2) fatigue of Type 316NG SS, and (3) SCC of ferritic steels used in reactor piping, pressure vessels, and steam generators. Stress corrosion cracking studies on austentitic SS explored the critical strains required for crack initiation, the effects of crevice conditions on SCC susceptibility, heat-to-heat variations in SCC susceptibility of Type 316NG and modified Type 347 SS, the effect of heat treatment on the susceptibility of Type 347 SS, threshold stress intensity values for crack growth in Type 316NG SS, and the effects of cuprous ion and several organic salts on the SCC of sensitized Type 304 SS. Crevice conditions were observed to strongly promote SCC. Significant heat-to-heat variations were observed in SCC susceptibility of Types 316NG and 347 SS. No correlation was found between SCC behavior and minor variations in chemical composition. A significant effect of heat treatment was observed in Type 347 SS. A heat that was extremely resistant to SCC after heat treatment at 650/degree/C for 24 h was susceptible to transgranular stress corrosion cracking (TGSCC) in the solution-annealed condition. Although there was no sensitization in either condition, the presence or absence of precipitates and differences in precipitate morphology appear to influence the SCC behavior. 20 refs., 20 figs., 11 tabs.
Date: October 1, 1988
Creator: Shack, W.J.; Kassner, T.F.; Maiya, P.S.; Park, J.Y. & Ruther, W.E.
Partner: UNT Libraries Government Documents Department

Sampling and Analysis Procedures for Gas, Condensate, Brine, and Solids: Pleasant Bayou Well Test, 1988-Present

Description: This section covers analyses performed on gas. Chemical analyses can only be related to well performance if the quantity of the various fluids are known. The IGT on-line data computer system measures the flowrate, the pressures, and the temperatures every 10 seconds. These values are automatically recorded over operator selected intervals both on magnetic media and on paper. This allows review of samples versus operating conditions. This paper covers analyses performed on gas, including: An approximate sampling schedule during flow tests; On-site sample handling and storage of gas samples; Addresses of laboratories that perform off site analyses; Sample shipping instructions; Data archiving; and Quality Control/Quality Assurance. It is expected that the above procedures will change as the flow test progresses, but deviations from the written procedures should be approved by C. Hayden of IGT and noted on the results of the analysis.
Date: January 1, 1988
Creator: Hayden, Chris
Partner: UNT Libraries Government Documents Department

Climate Change Reference Book

Description: This report was prepared for DOE Deputy Assistant Secretary for Renewable Energy, in 1988, by Meridian Corporation, Alexandria, Va. It set the environmental impacts of geothermal power in the context of those of other renewables, coal, and nuclear. While this is a compendium from largely secondary sources, it gives estimates for CO2 per GWh for the three phases of resource extraction, facility construction, and facility operation. Tons of concrete and steel used are estimated. System include hydrothermal (dry steam, flashed, binary), geopressured, hot dry rock, and magma. (DJE 2005)
Date: January 1, 1988
Partner: UNT Libraries Government Documents Department

Thermal-hydraulic response and iodine transport during a steam generator tube rupture

Description: Recent reanalyses of the offsite dose consequences following a steam generator tube rupture have identified a possible non-conservatism in original FSAR analyses. Post-trip uncovery of the top of the steam generator U-tubes, in conjunction with a break near the U-tube top, could lead to increased iodine release due to a reduced ''scrubbing'' of the iodine in the primary break fluid by the steam generator secondary liquid. To evaluate this issue, analyses were performed at the Idaho National Engineering Laboratory. The RELAP5 computer code was used to conduct an analysis of the Surry plant to determine whether the post-trip steam generator secondary mixture level was sufficient to maintain continuous coverage of the U-tubes. The results indicated continuous coverage of the U-tubes. The RELAP5 result was supported by a hand calculation. Additional RELAP5 analyses were conducted to determine magnitudes of iodine release for a steam generator tube rupture. Two sensitivity studies were conducted. The amount of iodine released to the atmosphere was strongly dependent on the assumed value of the partition coefficient. The assumption of steam generator U-tube uncovery, on a collapsed liquid level basis, following reactor trip had a minor effect on the amount of released iodine. 17 refs., 28 figs., 5 tabs.
Date: October 1, 1988
Creator: Callow, R.A.
Partner: UNT Libraries Government Documents Department

MHTGR: New production reactor summary of experience base

Description: Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs.
Date: March 1, 1988
Partner: UNT Libraries Government Documents Department

A flammability and combustion model for integrated accident analysis. [Advanced light water reactors]

Description: A model for flammability characteristics and combustion of hydrogen and carbon monoxide mixtures is presented for application to severe accident analysis of Advanced Light Water Reactors (ALWR's). Flammability of general mixtures for thermodynamic conditions anticipated during a severe accident is quantified with a new correlation technique applied to data for several fuel and inertant mixtures and using accepted methods for combining these data. Combustion behavior is quantified by a mechanistic model consisting of a continuity and momentum balance for the burned gases, and considering an uncertainty parameter to match the idealized process to experiment. Benchmarks against experiment demonstrate the validity of this approach for a single recommended value of the flame flux multiplier parameter. The models presented here are equally applicable to analysis of current LWR's. 21 refs., 16 figs., 6 tabs.
Date: January 1, 1988
Creator: Plys, M.G.; Astleford, R.D. & Epstein, M. (Fauske and Associates, Inc., Burr Ridge, IL (USA))
Partner: UNT Libraries Government Documents Department

Timing of the Three Mile Island Unit 2 core degradation as determined by forensic engineering

Description: Unlike computer simulation of an event, forensic engineering is the evaluation of recorded data and damaged as well as surviving components after an event to determine progressive causes of the event. Such an evaluation of the 1979 Three Mile Island Unit 2 accident indicates that gas began accumulating in steam, generator A at 6:10, or 130 min into the accident and, therefore, fuel cladding ruptures and/or zirconium-water reactions began at that time. Zirconium oxidation/hydrogen generation rates were highest ({approximately}70 kg of hydrogen per minute) during the core quench and collapse at 175 min. By 180 min, over 85% of the hydrogen generated by the zirconium-water reaction had been produced, and {approximately}400 kg of hydrogen had accumulated in the reactor coolant system. At that time, hydrogen concentrations at the steam/water interfaces in both steam generators approached 90%. By 203 min, the damaged reactor core had been reflooded and has not been uncovered since that time. Therefore, the core was completely under water at 225 min, when molten core material flowed into the lower head of the reactor vessel. 10 refs., 7 figs., 1 tab.
Date: January 1, 1988
Creator: Henrie, J.O. (Hydrogen Control, Inc., Panguitch, UT (USA))
Partner: UNT Libraries Government Documents Department

Aging assessment of PWR (Pressurized Water Reactor) Auxiliary Feedwater Systems

Description: In support of the Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, Oak Ridge National Laboratory is conducting a review of Pressurized Water Reactor Auxiliary Feedwater Systems. Two of the objectives of the NPAR Program are to identify failure modes and causes and identify methods to detect and track degradation. In Phase I of the Auxiliary Feedwater System study, a detailed review of system design and operating and surveillance practices at a reference plant is being conducted to determine failure modes and to provide an indication of the ability of current monitoring methods to detect system degradation. The extent to which current practices are contributing to aging and service wear related degradation is also being assessed. This paper provides a description of the study approach, examples of results, and some interim observations and conclusions. 1 fig., 1 tab.
Date: January 1, 1988
Creator: Casada, D.A.
Partner: UNT Libraries Government Documents Department