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Risk effectiveness evaluation of surveillance testing

Description: In nuclear power plants surveillance tests are required to detect failures in standby safety system components as a means of assuring their availability in case of an accident. However, the performance of surveillance tests at power may have adverse impact on safety as evidenced by the operating experience of the plants. The risk associated with a test includes two different aspects: (1) a positive aspect, i.e., risk contribution detected by the test, that results from the detection of failures which occur between tests and are detected by the test, and (2) a negative aspect, i.e., risk contribution caused by the test, that includes failures and degradations which are caused by the test or are related to the performance of the test. In terms of the two different risk contributions, the risk effectiveness of a test can be simply defined as follows: a test is risk effective if the risk contribution detected by the test is greater than the risk contribution caused by the test; otherwise it is risk ineffective. The methodology presentation will focus on two important kinds of negative test risk impacts, that is, the risk impacts of test-caused transients and equipment wear-out. The evaluation results of the risk effectiveness of the test will be presented in the full paper along with the risk assessment methodology and the insights from the sensitivity analysis. These constitute the core of the NUREG/CR-5775.
Date: July 20, 1992
Creator: Martorell, S.; Kim, I. S. (Universidad Politecnica de Valencia (Spain). Dept. de Ingenieria Quimica y Nuclear); Samanta, P. K. (Brookhaven National Lab., Upton, NY (United States)) & Vesely, W. E. (Science Applications International Corp., Columbus, OH (United States))
Partner: UNT Libraries Government Documents Department

Evaluation of sustained load effects in reactor pressure vessels by means of intermediate-scale flawed vessel tests. [BWR; PWR]

Description: Application of results of fracture tests of flawed vessels to full-scale reactor pressure vessels must take into account differences in scale, which may affect both the stress field around the flaw and the effective material toughness, as well as differences in the loading system which affect crack behavior through load variations. The intermediate test vessels (ITV's) of the Heavy Section Steel Technology program were designed with essentially full-scale thickness so that fracture initiation behavior could be demonstrated with minimal uncertainty associated with scale. In most of the ITV tests the stiffness of the hydraulic loading system relative to that of a real reactor system was of no consequence, since the terminal fracture was rapid and extensive. Two of the ITV tests resulted in leak without burst, however, which suggested the study of the behavior of intermediate test vessels under sustained load, as would be obtained in a large pressurized-water or boiling-water reactor system. Consequently the test of ITV-7 has been repeated with a flaw and test conditions nearly identical to the original flaw and conditions but with the load imposed pneumatically and the flawed region covered with a patch to retard or prevent leakage at the time of rupture. The results of the tests suggest that the demonstrations of leak without burst in the intermediate vessel tests, both hydraulic and pneumatic, are applicable to the evaluation of the behavior of reactor pressure vessels with similar flaw geometries under sustained load. The two vessels that ruptured in this way withstood pressures 2.15 to 2.74 times design pressure. These test pressures are above the ASME Boiler and Pressure Vessel Code allowable pressures for faulted conditions.
Date: January 1, 1976
Creator: Bryan, R. H.
Partner: UNT Libraries Government Documents Department

Method for determining the fuel contribution to the source term in transport casks

Description: Detailed models and analytical procedures are applied to the many complex aspects of spent fuel in transport including characterization of the fuel's irradiation conditions and initial states at the time of shipment, determination of the dynamic forces on the fuel assemblies that result from regulatory transportation accidents, modeling and analysis of the fuel's mechanical response to these forces, and estimation of the releasable radioactivity in the event of cladding breach. The methodology adopts a combined deterministic/probabilistic analysis approach in which each aspect of the problem is appropriately treated on the basis of its level of determinability. The results are obtained in the form of failure probabilities for each regulatory event considered. 3 refs., 6 figs.
Date: January 1, 1989
Creator: Rashid, Y. R.; Lake, W. H. & Sanders, T. L.
Partner: UNT Libraries Government Documents Department

Burst testing of alloys 800 and 310 at 1,255 K (1,800/sup 0/F) with a simulated coal gasification atmosphere

Description: Several corrosion- and heat-resistant alloys are being considered for long term applications in coal gasification plants at temperatures up to 1.255/sup 0/K in high pressure environments of mixed hydrogen, water, hydrocarbons, and sulfides. A method for in situ testing has been developed for short time mechanical tests of candidate alloys in high pressure, high temperature, gaseous environments, referred to as coal gasification atmosphere (CGA). The method involves bursting thin-walled tubes, using various gases to produce the burst hoop stress. The short time 1.255/sup 0/K burst and creep rupture strength and ductility properties of alloys 800 and 800H in a mixed gas environment, H/sub 2/, CO, CO/sub 2/, CH/sub 4/, SO/sub 2/ (CGA), are not reduced from properties obtained in air. However, the stress- and pressure-accelerated corrosion is more severe in CGA. It is expected that CGA will reduce long term strength and ductility in alloy 800 as a result of the accelerated corrosion. The short time 1.255/sup 0/K strengths of alloy 310 in CGA and pure hydrogen environments are reduced from the values obtained in air by less than 10 percent. The ductilities (total circumferential elongation) are good--approximately 20 percent for all test conditions. The CGA stress- and pressure-accelerated corrosion is greater than in air. Longer time tests in CGA are expected to result in additional strength degradation. Limited creep/fatigue tests of alloy 310 in hydrogen show that hold times are significant. A greater cyclic life is observed using an 8-second hold time than a 55-second hold time.
Date: May 1, 1976
Creator: Dixon, C. E.
Partner: UNT Libraries Government Documents Department

Preliminary Failure Modes, Effects and Criticality Analysis (FMECA) of the conceptual Brayton Isotope Power System (BIPS) Flight System

Description: A failure modes, effects and criticality analysis (FMECA) was made of the Brayton Isotope Power System Flight System (BIPS-FS) as presently conceived. The components analyzed include: Mini-BRU; Heat Source Assembly (HSA); Mini-Brayton Recuperator (MBR); Space Radiator; Ducts and Bellows, Insulation System; Controls; and Isotope Heat Source (IHS). (TFD)
Date: January 12, 1976
Creator: Miller, L. G.
Partner: UNT Libraries Government Documents Department

Joint U. S. --U. S. S. R. test of U. S. MHD electrode systems in U. S. S. R. U-02 MHD facility (phase I). Final report

Description: The first (Phase I) joint U.S.-U.S.S.R. test of U.S. electrode materials was carried out in Moscow between September 25 and October 8, 1975 in the Soviet U-02 MHD facility. The test procedure followed closely a predetermined work plan designed to test five different zirconia based materials and the cathode and anode electrode wall modules under MHD operating conditions. The materials which were selected were 88Zr0/sub 2/-12Y/sub 2/0/sub 3/, 82Zr0/sub 2/-18Ce02, 50Zr0/sub 2/-50Ce0/sub 2/, 25Zr0/sub 2/-75Ce0/sub 2/ and 20Zr0/sub 2/-78Ce0/sub 2/-2Ta/sub 2/0/sub 5/. The electrode modules were constructed by Westinghouse Research and Development Laboratory. Each of the five electrode materials had four different current densities established between the anode and cathode during the experiment which lasted a total of 127 hours. There were four main phases in the test schedule: (1) start-up of the channel over a specific heating period. No seed (K/sub 2/C0/sub 3/) introduction - 18 hours. (2) Electrical tests at operating temperature to investigate electro-physical characteristics of the channel and electrodes - 6 hours. (3) Operating life test - 94 hours. (4) Shut-down of the channel over a specific cool down period - 9 hours. All except six electrode pairs performed satisfactorily during the entire test. These were the pairs which were designated to carry maximum or near maximum current density. Five pairs failed early in the life test and the sixth pair failed in the last several hours. Failure was not due to the electrode materials, however, but due to lead-out melting caused by joule heating in the platinum wires. The U-02 facility is described and the operational parameters are given for each phase of the test. The electrode and insulating walls are described and the appropriate parameters that are used to predict the performance of the module are given.
Date: January 1, 1976
Creator: Hosler, W R
Partner: UNT Libraries Government Documents Department

Status report on the MHF mapping and characterization program

Description: The surface electrical potential system was refurbished and updated during this year prior to conducting several field experiments. Results from these MHF's not only include fracture orientation but also provided some insight into fracture growth periods. A very shallow fracture experiment was also conducted to calibrate the models and allowed verification of several different mapping techniques. The surface seismic recording effort has been terminated and its emphasis placed on downhole, wall clamped, three-axis geophone recording system. This system should be available for testing during fiscal '78.
Date: January 1, 1977
Creator: Schuster, C. L.
Partner: UNT Libraries Government Documents Department

Reliability of CRBR primary piping: critique of stress-strength overlap method for cold-leg inlet downcomer

Description: A critique is presented of the strength-stress overlap method for the reliability of the CRBR primary heat transport system piping. The report addresses, in particular, the reliability assessment of WARD-D-0127 (Piping Integrity Status Report), which is part of the CRBR PSAR docket. It was found that the reliability assessment is extremely sensitive to the assumed shape for the probability density function for the strength (regarded as a random variable) of the cold-leg inlet downcomer section of the primary piping. Based on the rigorous Chebyschev inequality, it is shown that the piping failure probability is less than 10/sup -2/. On the other hand, it is shown that the failure probability can be much larger than approximately 10/sup -13/, the typical value put forth in WARD-D-0127.
Date: April 1, 1976
Creator: Bari, R. A.; Buslik, A. J. & Papazoglou, I. A.
Partner: UNT Libraries Government Documents Department

Effluent and sanitary sewer monitors

Description: Two similar instruments that monitor the liquid wastes from the plutonium facility are described. The operation of the two instruments is completely automatic and performs a continuous surveillance in the frame of Nuclear Safeguards. One instrument controls the liquids from the facility and the other checks the sanitary sewer wastes. Both have self-diagnosing capabilities and take automatic actions in case of abnormal occurrences.
Date: March 1, 1977
Creator: Stanchi, L. & Vasey, M. R.
Partner: UNT Libraries Government Documents Department

Reliability and operating experience of the LAMPF 805-MHz rf system

Description: Over 850,000 hours have been accumulated on the klystrons and modulators that constitute the LAMPF 805-MHz rf system. The 1-1/4 MW klystrons, the floating-deck modulators, and the modulator triodes are described. The operating data are summarized, and the fault and failure modes are tabulated for the three major components of the system. The high-voltage processing and other maintenance required to keep this 86-kV system operating reliably are described. The tube failure rates, tube fault rates, and modulator fault rates are presented. The mean time to failures is greater than 56,500 h for the klystrons and greater than 40,600 h for the modulator triodes. The steps taken to produce such good reliability are discussed.
Date: January 1, 1976
Creator: Tallerico, P. J.
Partner: UNT Libraries Government Documents Department

Investigation of fracture in pressurized gas metal arc welded beryllium

Description: Premature failures during proof testing of pressurized-gas-metal-arc (PGMA) welded beryllium assemblies were investigated. The failures were almost entirely within the beryllium (a forming grade, similar to HP-10 or S-240), close to and parallel to the weld interface. The aluminum-silicon weld filler metal deposit was not centered in the weld groove in the failed assemblies, and failure occurred on the side of the weld opposite the bias in the weld deposit. Tensile tests of welded samples demonstrated that the failures were unrelated to residual machining damage from cutting the weld groove, and indicated small lack-of-fusion areas near the weld start to be the most likely origin of the failures. Acoustic emission was monitored during tensile tests of the welds. The majority of acoustic emission was probably from crack propagation through the weld filler metal. Tensile bars cut from the region of the weld start behaved differently; they failed at lower loads and exhibited an acoustic emission behavior believed to be from cracking in the weld metal-beryllium interface. Improvement in the quality of these and similar beryllium welds can therefore most likely be made by centering the weld deposit and reducing the size of the weld start defect. 21 fig.
Date: May 20, 1976
Creator: Heiple, C. R.; Merlini, R. J. & Adams, R. O.
Partner: UNT Libraries Government Documents Department

An evaluation of LMR design options for reduction of sodium void worth

Description: In this study, we analyze the relationship between the sodium void worth ({rho}{sub NA}) and other important performance characteristics for various design options which reduce {rho}{sub NA}. Our objective was to identify a preferred design option for reducing {rho}{sub NA} based on an overall consideration of performance tradeoffs. The focus of this study is on core designs of recent interest in the US LMR program, i.e. designs in the 450 to 1200 MWt size range that make use of metal alloy fuel. A key objective of the LMR development program in the US has been to design cores that can passively avoid damage when the control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). Analyses and experimental tests of such unprotected events have demonstrated that the physical properties of metallic fuel alloys and the neutronic feedback characteristics of metal-fueled cores can be exploited to obtain favorable relations among the power, power/flow, and inlet temperature coefficients of reactivity and, consequently, large margins to sodium boiling and fuel damage under accident conditions. Since the reactivity effects of sodium density variation during postulated accidents are effectively compensated by other feedback effects, reduction of the sodium void worth has not been a primary design objective for recent LMR concepts; relatively large values ($4 to $6) are predicted for current core designs. 23 refs., 11 figs.
Date: January 1, 1989
Creator: Hill, R N & Khalil, H
Partner: UNT Libraries Government Documents Department

Fluid transport properties of rock fractures at high pressure and temperature. Progress report, July 1, 1976--June 30, 1977

Description: Initial stages of a study on the fluid transport properties of rock at high pressure and temperature are reported. Emphasis is placed on the mechanical hydraulic interactions, in an attempt to clarify the process of fracture closure and its influence on fracture permeability. To determine the fluid transport properties of a fracture, the effect of surface roughness, geometry, and filling on fracture permeability was investigated. Permeability of these fractures was measured at various effective normal stresses at room temperature. The law of effective stress appears valid for fractures without filling but permeability of filled fractures is more sensitive to confining pressure than pore pressure. Permeability of smooth surfaces varied 5 to 0.5 darcys over a range of effective stresses from 0 to 3000 bars. Filled fractures were an order of magnitude more permeable.
Date: March 1, 1977
Creator: Engelder, T. & Scholz, C.
Partner: UNT Libraries Government Documents Department

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

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

Effect of sodium on the creep-rupture behavior of type 304 stainless steel

Description: Uniaxial creep-rupture data have been obtained for Type 304 stainless steel in the solution-annealed condition and after exposure to a flowing sodium environment at temperatures of 700, 650, and 600/sup 0/C.The specimens were exposed to sodium for time periods between 120 and 5012 h to produce carbon penetration depths of approximately 0.010, 0.020, and 0.038 cm in the steel. Results showed that, as the depth of carbon penetration and the average carbon concentration in the steel increase, the rupture life increases and the minimum creep rate decreases. Creep correlations that relate rupture life, minimum creep rate, and time-to-tertiary creep were developed for the steel in both the solution-annealed and sodium-exposed conditions. Isochronous stress-creep strain curves and results on the calculations of the stress levels for 1 percent creep strain and long-term rupture life are also presented. 11 fig.
Date: January 1, 1976
Creator: Natesan, K.; Chopra, O. K. & Kassner, T. F.
Partner: UNT Libraries Government Documents Department

Interpretation of the pressure and flow data for the two fractures of the Los Alamos hot dry rock (HDR) geothermal system

Description: Since the initial hydraulic fractures were established at the Fenton Hill site of the Los Alamos HDR project, several pressurization and flow experiments have been performed. The fractures are in granite at a depth of approximately 2900 meters and are separated by approximately 30 meters. The flow experiments were planned to establish the water losses to the surrounding rock, determine pressure and stress dependent rock properties, and to characterize the fracture system in terms of extent, volume, and the interconnection between the fractures. Here, an analysis of these experiments is presented in terms of a mathematical model that includes the variable rock permeability and porosity, the connection between permeability and porosity as given by simple models, and the effects of the earth stresses. These features are incorporated into the diffusion equation for the pore pressure. The experimental relationship of pressure and flow in the two fractures is examined.
Date: January 1, 1977
Creator: Fisher, H. N.
Partner: UNT Libraries Government Documents Department

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

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

Magnetic fusion energy materials technology program, annual progress report for period ending June 30, 1976

Description: Activities in research programs are reported on materials for use in thermonuclear reactor development. Information and data are included on radiation effects on stainless steel 316, nickel-base alloys, molybdenum-base alloys, vanadium alloys, and SAP. Results of compatibility studies involving iron-base alloys and lithium are also included along with research results on magnet development. (JRD)
Date: September 1, 1976
Creator: Scott, J. L. (comp.)
Partner: UNT Libraries Government Documents Department

Development of high-temperature acoustic instrumentation for characterization of hydraulic fractures in dry hot rock. [Downhole geophone sonde]

Description: The primary objectives of the post hydraulic fracture experiments in Geothermal Test Hole No. 2 are to study methods of measuring the location, orientation, and shape of the crack and to determine the stability of pressurized fracture systems. Detection of fracture dimensions and orientation of the geothermal reservoir is important for creating and understanding the operation of a dry hot rock energy-extraction system. These objectives require development of downhole instrumentation capable of characterization of hydraulic-fracture systems in high-temperature and high-pressure borehole environments. The development of the downhole instrumentation must emphasize reliability of measuring devices and electromechanical components to function properly at borehole temperatures of 250/sup 0/C and pressures of 690 bars (10,000 psi).
Date: January 1, 1976
Creator: Dennis, B. R.; Hill, J. H.; Stephani, E. L. & Todd, B. E.
Partner: UNT Libraries Government Documents Department

Annotated bibliography of safety-related occurrences in pressurized-water nuclear power plants as reported in 1975

Description: The bibliography presented contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at pressurized-water reactor nuclear power plants in 1975. The report includes 1097 abstracts, arranged alphabetically by reactor name and then chronologically for each reactor, that describe incidents, failures, and design or construction deficiencies experienced at the facilities. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables summarizing the information contained in the bibliography are presented. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). A few of the unique events that occurred during the year are reviewed in detail.
Date: July 1, 1976
Creator: Scott, R. L. & Gallaher, R. B.
Partner: UNT Libraries Government Documents Department

Actuator system history of safety rod lower latch problems review of latch inspection video tapes

Description: During pre-restart testing the safety rod at position X26-YlO bound after being driven approximately two (2) feet out of the reactor. Subsequently, the rod was manually returned to it's seated position. Inspection of the lower latch showed that the latch locking plunger button (screwed on to the bottom of the plunger shaft and retained by a pin through a hole drilled through the button and the plunger shaft) was missing. The shaft failed through the hole drilled for the retaining pin. The button, with the retaining pin intact, was found lodged between the safety rod upper adapter collar and the top of the safety rod thimble top fitting. Analysis of the safety rod latch and accompanying forest guide tube design provided assurance that this type of failure would not cause binding during the scramming'' of the safety rods. Inspection of all of the K'' safety rod lower latches revealed six other latches with missing plunger buttons, and nine with other non-conformances which required latch replacement. A history search conducted by Reactor Engineering Design, Components Handling Group, is included in this report. The history search shows that latch design modifications, as a part of initial development of the latch system and later to improve the delatching operation, were made from 1950 to 1960. These modifications created a condition where latch damage could occur. Video tapes were made during inspection of the safety rod latches in K area and control rod latches in L area. These tapes were reviewed by Reactor Engineering Design Components Handling engineers. The reviews were used for correlation of latch problems reported by the engineers/mechanics making the inspections. The K area tapes showed inspection of 65 of the 66 safety rod latches. The review of the tapes showed the plunger buttons to be missing from five latches. RED-CH ...
Date: June 24, 1992
Creator: Banks, J. J.
Partner: UNT Libraries Government Documents Department

High-temperature deformation and rupture behavior of internally-pressurized Zircaloy-4 cladding in vacuum and steam enivronments. [LOCA conditions]

Description: The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant accident situations in light-water reactors. The effects of internal pressure, heating rate, axial constraint, and localized temperature nonuniformities in the cladding on the maximum circumferential strain have been determined for burst temperatures between approximately 650 and 1350/sup 0/C.
Date: January 1, 1977
Creator: Chung, H. M.; Garde, A. M. & Kassner, T. F.
Partner: UNT Libraries Government Documents Department