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Laser ablation of concrete.

Description: Laser ablation is effective both as an analytical tool and as a means of removing surface coatings. The elemental composition of surfaces can be determined by either mass spectrometry or atomic emission spectroscopy of the atomized effluent. Paint can be removed from aircraft without damage to the underlying aluminum substrate, and environmentally damaged buildings and sculptures can be restored by ablating away deposited grime. A recent application of laser ablation is the removal of radioactive contaminants from the surface and near-surface regions of concrete. We present the results of ablation tests on concrete samples using a high power pulsed Nd:YAG laser with fiber optic beam delivery. The laser-surface interaction was studied on various model systems consisting of Type I Portland cement with varying amounts of either fine silica or sand in an effort to understand the effect of substrate composition on ablation rates and mechanisms. A sample of non-contaminated concrete from a nuclear power plant was also studied. In addition, cement and concrete samples were doped with non-radioactive isotopes of elements representative of cooling waterspills, such as cesium and strontium, and analyzed by laser-resorption mass spectrometry to determine the contamination pathways. These samples were also ablated at high power to determine the efficiency with which surface contaminants are removed and captured. The results show that the neat cement matrix melts and vaporizes when little or no sand or aggregate is present. Surface flows of liquid material are readily apparent on the ablated surface and the captured aerosol takes the form of glassy beads up to a few tens of microns in diameter. The presence of sand and aggregate particles causes the material to disaggregate on ablation, with intact particles on the millimeter size scale leaving the surface. Laser resorption mass spectrometric analysis showed that cesium and potassium have similar chemical ...
Date: October 5, 1998
Creator: Savina, M.
Partner: UNT Libraries Government Documents Department

A Radiological Survey Approach to Use Prior to Decommissioning: Results from a Technology Scanning and Assessment Project Focused on the Chornobyl NPP

Description: The primary objectives of this project are to learn how to plan and execute the Technology Scanning and Assessment (TSA) approach by conducting a project and to be able to provide the approach as a capability to the Chernobyl Nuclear Power Plant (ChNPP) and potentially elsewhere. A secondary objective is to learn specifics about decommissioning and in particular about radiological surveying to be performed prior to decommissioning to help ChNPP decision makers. TSA is a multi-faceted capability that monitors and analyzes scientific, technical, regulatory, and business factors and trends for decision makers and company leaders. It is a management tool where information is systematically gathered, analyzed, and used in business planning and decision making. It helps managers by organizing the flow of critical information and provides managers with information they can act upon. The focus of this TSA project is on radiological surveying with the target being ChNPP's Unit 1. This reactor was stopped on November 30, 1996. At this time, Ukraine failed to have a regulatory basis to provide guidelines for nuclear site decommissioning. This situation has not changed as of today. A number of documents have been prepared to become a basis for a combined study of the ChNPP Unit 1 from the engineering and radiological perspectives. The results of such a study are expected to be used when a detailed decommissioning plan is created.
Date: October 20, 1999
Creator: Milchikov, A.; Hund, G. & Davidko, M.
Partner: UNT Libraries Government Documents Department

Comparisons of HELIOS Calculated Isotope Concentrations to Measured Values for Several Reactor Systems

Description: Heavy metal and fission product noble gas concentrations in spent fuel from two different PWR'S were calculated using HELIOS and compared to measured results from the literature. It was found that for the U-235/U-238 and Pu-240/Pu-239 isotopic ratios, the HELIOS calculation agreed to within the experimental uncertainty. For the Xe-131/Xe-134 isotopic ratios, HELIOS tended to overestimate the result by up to 4%. Conversely for the Xe-132/Xe-134 ratios, HELIOS underestimated the result by a slight amount ({approximately}1%). This suggests that either the fission product yields for Xe-131 and Xe-132 should be slightly altered or that the absorption cross-section for Xe-131 should be slightly increased. More analysis is necessary to determine which of these two alternatives is more appropriate. This work has shown that the accuracy of HELIOS (within 2% for heavy metals and within 4% for fission noble gases) is sufficient for most analyses.
Date: October 21, 1998
Creator: Charlton, W.S.; Perry, R.T.; Fearey, B.L. & Parish, T.A.
Partner: UNT Libraries Government Documents Department

Synergistic failure of BWR internals

Description: Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components.
Date: October 25, 1999
Creator: Ware, A. G. & Chang, T. Y.
Partner: UNT Libraries Government Documents Department

The corrosion behavior of hafnium in high-temperature-water environments

Description: The high-temperature-water corrosion performance of hafnium is evaluated. Corrosion kinetic data are used to develop correlations that are a function of time and temperature. The evaluation is based on corrosion tests conducted in out-of-pile autoclaves and in out-of-flux locations of the Advanced Test Reactor (ATR) at temperatures ranging from 288 to 360 C. Similar to the corrosion behavior of unalloyed zirconium, the high-temperature-water corrosion response of hafnium exhibits three corrosion regimes: pretransition, posttransition, and spalling. In the pretransition regime, cubic corrosion kinetics are exhibited, whereas in the posttransition regime, linear corrosion kinetics are exhibited. Because of the scatter in the spalling regime data, it is not reasonable to use a best fit of the data to describe spalling regime corrosion. Data also show that neutron irradiation does not alter the corrosion performance of hafnium. Finally, the data illustrate that the corrosion rate of hafnium is significantly less than that of Zircaloy-2 and Zircaloy-4.
Date: October 1, 1999
Creator: Rishel, D.M.; Smee, J.D. & Kammenzind, B.F.
Partner: UNT Libraries Government Documents Department

Estimating the Uncertainty in Reactivity Accident Neutronic Calculations

Description: A study of the uncertainty in calculations of the rod ejection accident in a pressurized water reactor is being carried out for the US Nuclear Regulatory Commission. This paper is a progress report on that study. Results are presented for the sensitivity of core energy deposition to the key parameters: ejected rod worth, delayed neutron fraction, Doppler reactivity coefficient, and fuel specific heat. These results can be used in the future to estimate the uncertainty in local fuel enthalpy given some assumptions about the uncertainty in the key parameters. This study is also concerned with the effect of the intra-assembly representation in calculations. The issue is the error that might be present if assembly-average power is calculated, and pin peaking factors from a static calculation are then used to determine local fuel enthalpy. This is being studied with the help of a collaborative effort with Russian and French analysts who are using codes with different intra-assembly representations. The US code being used is PARCS which calculates power on an assembly-average basis. The Russian code being used is BARS which calculates power for individual fuel pins using a heterogeneous representation based on a Green's Function method.
Date: October 26, 1998
Creator: Diamond, D. J.; Yang, C. Y. & Aronson, A. L.
Partner: UNT Libraries Government Documents Department

Theory and application of deterministic multidimensional pointwise energy lattice physics method

Description: The theory and application of deterministic, multidimensional, pointwise energy lattice physics methods are discussed. These methods may be used to solve the neutron transport equation in multidimensional geometries using near-continuous energy detail to calculate equivalent few-group diffusion theory constants that rigorously account for spatial and spectral self-shielding effects. A dual energy resolution slowing down algorithm is described which reduces the computer memory and disk storage requirements for the slowing down calculation. Results are presented for a 2D BWR pin cell depletion benchmark problem.
Date: October 5, 1999
Creator: Zerkle, M.L.
Partner: UNT Libraries Government Documents Department

Sodium removal process development for LMFBR fuel subassemblies

Description: Two 37-pin scale models of Clinch River Breeder Reactor Plant fuel subassemblies were designed, fabricated and used at Westinghouse Advanced Reactors Division in the development and proof-testing of a rapid water-based sodium removal process for the ORNL Hot Experimental Facility, Liquid Metal Fast Breeder Reactor Fuel Reprocessing Cycle. Through a series of development tests on one of the models, including five (5) sodium wettings and three (3) high temperature sodium removal operations, optimum process parameters for a rapid water vapor-argon-water rinse process were identified and successfully proof-tested on a second model containing argon-pressurized, sodium-corroded model fuel pins simulating the gas plenum and cladding conditions expected for spent fuel pins in full scale subassemblies. Based on extrapolations of model proof test data, preliminary process parameters for a water vapor-nitrogen-water rinse process were calculated and recommended for use in processing full scale fuel subassemblies in the Sodium Removal Facility of the Fuel Receiving Cell, ORNL HEF.
Date: October 1, 1981
Creator: Simmons, C.R. & Taylor, G.R.
Partner: UNT Libraries Government Documents Department

Economic Study of Spent Nuclear Fuel Storage and Reprocessing Practices in Russia

Description: This report describes a study of nuclear power economics in Russia. It addresses political and institutional background factors which constrain Russia's energy choices in the short and intermediate run. In the approach developed here, political and institutional factors might dominate short-term decisions, but the comparative costs of Russia's fuel-cycle options are likely to constrain her long-term energy strategy. To this end, the authors have also formulated a set of policy questions which should be addressed using a quantitative decision modeling which analyzes economic costs for all major components of different fuel cycle options, including the evolution of uranium prices.
Date: October 1, 1997
Creator: Singer, C. E. & Miley, G. H.
Partner: UNT Libraries Government Documents Department

Final report of comprehensive testing program for concrete at elevated temperatures

Description: The objective of this program was to define the variations in physical (thermal) and mechanical (strength) properties of limestone aggregate concrete and lightweight insulating concrete exposed to elevated temperatures that could occur as a result of a postulated large sodium spill in a lined LMFBR equipment cell. To meet this objective, five test series were conducted: (1) unconfined compression, (2) shear, (3) rebar bond, (4) sustained loading (creep), and (5) thermal properties. Mechanical property results are presented for concretes subjected to temperature up to 621{sup 0}C (1150{sup 0}F).
Date: October 1, 1980
Creator: Oland, C.B.; Naus, D.J. & Robinson, G.C.
Partner: UNT Libraries Government Documents Department

Intercomparison of Results for a Pwr Rod Ejection Accident

Description: This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.
Date: October 1, 1999
Creator: Diamond, D. J.; Aronson, A.; Jo, J.; Avvakumov, A.; Malofeev, V.; Sidorov, V. et al.
Partner: UNT Libraries Government Documents Department

NRC Support for the Kalinin (Vver) Probabilistic Risk Assessment

Description: The US Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation have been working together since 1994 to carry out a probabilistic risk assessment (PRA) of a VVER-1000 in the Russian Federation. This was a recognition by both parties that this technology has had a profound effect on the discipline of nuclear reactor safety in the West and that the technology should be transferred to others so that it can be applied to Soviet-designed plants. The NRC provided funds from the Agency for International Development and technical support primarily through Brookhaven National Laboratory and its subcontractors. The latter support was carried out through workshops, by documenting the methodology to be used in a set of guides, and through periodic review of the technical activity. The result of this effort to date includes a set of procedure guides, a draft final report on the Level 1 PRA for internal events (excluding internal fires and floods), and progress reports on the fire, flood, and seismic analysis. It is the authors belief that the type of assistance provided by the NRC has been instrumental in assuring a quality product and transferring important technology for use by regulators and operators of Soviet-designed reactors. After a thorough review, the report will be finalized, lessons learned will be applied in the regulatory and operational regimes in the Russian Federation, and consideration will be given to supporting a containment analysis in order to complete a simplified Level 2 PRA.
Date: October 26, 1998
Creator: Bley, D.; Diamond, D. J.; Chu, T. L.; Azarm, A.; Pratt, W. T.; Johnson, D. et al.
Partner: UNT Libraries Government Documents Department

Chemical and Radiochemical Constituents in Water from Wells in the Vicinity of the Naval Reactors Facility, Idaho National Engineering and Environmental Laboratory, Idaho, 1996

Description: The U.S. Geological Survey, in response to a request from the U.S. Department of Energy's Pittsburgh Naval Reactors Office, Idaho Branch Office (IBO), samples water from 13 wells during 1996 as part of a long-term project to monitor water quality to the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility (NRF), Idaho National Engineering and Environmental Laboratory, Idaho. The IBO requires information about the mobility of radionuclide- and chemical-waste constituents in the Snake River Plain aquifer. Waste-constituent mobility is determined principally by (1) the rate and direction of ground-water flow; (2) the locations, quantities, and methods of waste disposal; (3) waste-constituents chemistry; and (4) the geochemical processes taking place in the aquifer. The purpose of the data-collection program is to provide IBO with water-chemistry data to evaluate the effect of NRF activities on the water quality of the Snake River Plain aquifer. Water samples were analyzed for naturally occurring constituents and man-made contaminants.
Date: October 1, 1999
Creator: Knobel, L. L.; Bartholomay, R. C.; Tucker, B. J. & Williams, L. M.
Partner: UNT Libraries Government Documents Department

Effects of Water Radiolysis in Water Cooled Reactors - Nuclear Energy Research Initiative (NERI) Program

Description: OAK B188 Quarterly Progress Report on NERI Proposal No.99-0010 for the Development of an Experiment and Calculation Based Model to Describe the Effects of Radiation on Non-standard Aqueous Systems Like Those Encountered in the Advanced Light Water Reactor
Date: October 1, 2000
Creator: Pimblott, S. M.
Partner: UNT Libraries Government Documents Department

Accuracy of the Quasistatic Method for Two-Dimensional Thermal Reactor Transients with Feedback

Description: An important aspect in the design and safe operation of a nuclear reactor is the behavior of a reactor in a transient, or nonsteady state, condition. This study shows that the quasistatic method is capable of producing highly accurate results, relative to the direct finite-difference method, for two-dimensional thermal reactor transients with feedback.
Date: October 23, 2001
Creator: Dodds, H.L. Jr.
Partner: UNT Libraries Government Documents Department

TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS.

Description: In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and subsequent accumulation of debris on the ...
Date: October 1, 2000
Creator: MAJI, A. K.; MARSHALL, B. & AL, ET
Partner: UNT Libraries Government Documents Department

Closed ThUOX Fuel Cycle for LWRs with ADTT (ATW) Backend for the 21st Century

Description: A future nuclear energy scenario with a closed, thorium-uranium-oxide (ThUOX) fuel cycle and new light water reactors (TULWRs) supported by Accelerator Transmutation of Waste (ATW) systems could provide several improvements beyond today's once-through, UO{sub 2}-fueled nuclear technology. A deployment scenario with TULWRs plus ATWs to burn the actinides produced by these LWRs and to close the back-end of the ThUOX fuel cycle was modeled to satisfy a US demand that increases linearly from 80 GWe in 2020 to 200 GWe by 2100. During the first 20 years of the scenario (2000-2020), nuclear energy production in the US declines from today's 100 GWe to about 80 GWe, in accordance with forecasts of the US DOE's Energy Information Administration. No new nuclear systems are added during this declining nuclear energy period, and all existing LWRs are shut down by 2045. Beginning in 2020, ATWs that transmute the actinides from existing LWRs are deployed, along with TULWRs and additional ATWs with a support ratio of 1 ATW to 7 TULWRs to meet the energy demand scenario. A final mix of 174 GWe from TULWRs and 26 GWe from ATWs provides the 200 GWe demand in 2100. Compared to a once-through LWR scenario that meets the same energy demand, the TULWR/ATW concept could result in the following improvements: depletion of natural uranium resources would be reduced by 50%; inventories of Pu which may result in weapons proliferation will be reduced in quantity by more than 98% and in quality because of higher neutron emissions and 50 times the alpha-decay heating of weapons-grade plutonium; actinides (and possibly fission products) for final disposal in nuclear waste would be substantially reduced; and the cost of fuel and the fuel cycle may be 20-30% less than the once-through UO{sub 2} fuel cycle.
Date: October 6, 1998
Creator: Beller, D.E.; Sailor, W.C. & Venneri, F.
Partner: UNT Libraries Government Documents Department

YOUR DESIGN PROBABLY NEEDS MORE VDUs

Description: The most frequent complaint of operators in modern computer-based control rooms is that there just are not enough video display units (VDUs). In this paper we examine the basis for this concern and try to understand the technical and historical reasons for this complaint, and its implications for the design of complex human-machine systems, including the number of VDUs in the control room. The overall aim of our work is to develop human factors guidance for the review of computer-based and modernized control rooms in nuclear power plants. As part of these efforts we have conducted literature reviews and studies using both simulators and actual systems in a broad range of industries, including process control, aerospace, medical, and others. Our findings reflect the general complaint of operators across all these industries: there just are not enough VDUs in the control room. We conclude that there are three primary reasons for this complaint. First, as part of a workload management strategy, operators frequently avoid interface management tasks and do not access all the information available, preferring instead to use a fixed set of familiar displays that provide much (but not all) of the information needed. Performance thereby becomes data limited and operators complain that they do not have a sufficient number of VDUs to set up in the early phases of a high-workload period so they can get all the information they need. Second, display designs are typically not designed with operator tasks in mind. The most common method of representing information is by functions and systems. Since tasks typically cut across many systems, operators need many displays. Thus, to make task performance easier operators need additional VDUs. Finally, there is a differing ''concept of operations'' between designers and operators. Modern computer-based control rooms are designed with vast amounts of data, ...
Date: October 8, 2001
Creator: OHARA, J.; BROWN, W.; LEWIS, P. & PERSENSKY, J.
Partner: UNT Libraries Government Documents Department