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Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

Description: This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).
Date: December 17, 2002
Creator: Wagner, J.C.
Partner: UNT Libraries Government Documents Department

Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

Description: This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.
Date: October 23, 2002
Creator: Wagner, J.C.
Partner: UNT Libraries Government Documents Department

Initial tests of thermoacoustic space power engine.

Description: Future NASA deep-space missions will require radioisotope-powered electric generators that are just as reliable as current RTGs, but more efficient and of higher specific power (Wikg). Thennoacoustic engines at the -1-kW scale have converted high-temperature heat into acoustic, or PV, power without moving parts at 30% efficiency. Consisting of only tubes and a few heat exchangers, thennoacoustic engines are low mass and promise to be highly reliable. Coupling a thennoacoustic engine to a low mass, highly reliable and efficient linear alternator will create a heat-driven electric generator suitable for deep-space applications. Conversion efficiency data will be presented on a demonstration thennoacoustic engine designed for the 1 00-Watt power range.
Date: January 1, 2002
Creator: Backhaus, S. N. (Scott N.)
Partner: UNT Libraries Government Documents Department

Spall experiments in convergent geometry using the atlas pulsed power facility.

Description: {sm_bullet}Four spall experiments have been performed using Atlas {sm_bullet} Purpose was to investigate damage in convergent geometry {sm_bullet} Impact pressures ranged between 45 kbars - 110 kbars {sm_bullet} Diagnostics included VISAR and axial and radial radiographs {sm_bullet} Targets were recovered for post-metallugical analysis
Date: January 1, 2002
Creator: Keinigs, R. K. (Rhonald K.); Anderson, W. A. (William A.); Cerreta, E. K. (Ellen K.); Cochrane, J. C. (James C.), Jr.; Ladish, J. S. (Joseph S.); Lindemuth, I. R. (Irvin R.) et al.
Partner: UNT Libraries Government Documents Department

Waste generation process modeling and analysis for fuel reprocessing technologies

Description: Estimates of electric power generation requirements for the next century, even when taking the most conservative tack, indicate that the United States will have to increase its production capacity significantly. If the country determines that nuclear power will not be a significant component of this production capacity, the nuclear industry will have to die, as maintaining a small nuclear component will not be justifiable. However, if nuclear power is to be a significant component, it will probably require some form of reprocessing technology. The once-through fuel cycle is only feasible for a relatively small number of nuclear power plants. If we are maintaining several hundred reactors, the once-through fuel cycle is more expensive and ethically questionable.
Date: January 1, 2002
Creator: Kornreich, D. E. (Drew E.); Koehler, A. C. (Andrew C.) & Farman, Richard F.
Partner: UNT Libraries Government Documents Department

A New Computational Tool for Simulation of 3-D Flow and Heat Transfer in Boiling Water Reactors

Description: This Phase I work has developed a novel hybrid Lattice Boltzmann Model for the simulation of nonideal fluid thermal dynamics and demonstrated that this model can be used to simulate fundamental two-phase flow processes including boiling initiation, bubble formation and coalescency, and flow-regime formation.
Date: December 9, 2002
Creator: Chen, Hudong
Partner: UNT Libraries Government Documents Department

Nonlinear Dynamics and Control of Large Arrays of Coupled Oscillators: Application to Fluid-Elastic Problems

Description: Large numbers of fluid elastic structures are part of many power plant systems and vibration of these systems sometimes are responsible for plant shut downs. Earlier research at Cornell in this area had centered on nonlinear dynamics of fluid-elastic systems with low degrees of freedom. The focus of current research is the study of the dynamics of thousands of closely arrayed structures in a cross flow under both fluid and impact forces. This research is relevant to two areas: (1) First, fluid-structural problems continue to be important in the power industry, especially in heat exchange systems where up to thousands of pipe-like structures interact with a fluid medium. [Three years ago in Japan for example, there was a shut down of the Monju nuclear power plant due to a failure attributed to flow induced vibrations.] (2) The second area of relevance is to nonlinear systems and complexity phenomena; issues such as spatial temporal dynamics, localization and coherent patterns entropy measures as well as other complexity issues. Early research on flow induced vibrations in tube row and array structures in cross flow goes back to Roberts in 1966 and Connors in 1970. These studies used linear models as have many of the later work in the 1980's. Nonlinear studies of cross flow induced vibrations have been undertaken in the last decade. The research at Cornell sponsored by DOE has explored nonlinear phenomena in fluid-structure problems. In the work at Cornell we have documented a subcritical Hopf bifurcation for flow around a single row of flexible tubes and have developed an analytical model based on nonlinear system identification techniques. (Thothadri, 1998, Thothadri and Moon, 1998, 1999). These techniques have been applied to a wind tunnel experiment with a row of seven cylinders in a cross flow. These system identification methods have been ...
Date: April 1, 2002
Creator: Moon, Francis C.
Partner: UNT Libraries Government Documents Department

Study of safeguards system on dry reprocessing for fast breeder reactor

Description: A 'Feasibility Study on the Commercialized Fast Breeder Reactor (FBR) Cycle System' is underway at Japan Nuclear Cycle Development Institute (JNC). Concepts to commercialize the FBR fuel cycle are being created together with their necessary research and development (R&D) tasks. 'Dry,' non-aqueous, processes are candidates for FBR fuel reprocessing. Dry reprocessing technology takes advantage of proliferation barriers, due to the lower decontamination factors achievable by the simple pyrochemical processes proposed. The concentration o f highly radioactive impurities and non-fissile materials in products from a dry reprocess is generally significantly larger than the normal aqueous (Purex) process. However, the safeguards of dry reprocesses have not been widely analyzed. In 2000, JNC and Los Alamos National Laboratoiy (LANL) initiated a joint research program to study the safeguards aspects of dry reprocessing. In this study, the safeguardability of the three options: metal electrorefining, oxide electrowinning, and fluoride volatility processes, are assessed. FBR spent fuels are decladded and powdered into mixed oxides (MOX) at the Head-End process either by oxidation-reduction reactions (metal electrorefining and fluoride volatility) or mechanically (oxide electrowinning). At the oxide electrowinning process, the spent MOX he1 powder is transferred to chloride in molten salt and nuclear materials are extracted onto cathode as oxides. For metal electrorefining process, on the other hand, the MOX fuel is converted to chloride in molten salt, and nuclear materials are extracted onto cathode as a metal fomi. At lhe fluoride volatility process, the MOX fuel powder is converted to U&/PuF6 (gaseous form) in a fluidized bed; plutonium and uranium fluorides are separated by volatilization properties and then are converted to oxides. Since the conceptual design of a dry reprocessing plant is incomplete, the operational mode, vessel capacities, residence times, and campaigns are not fully defined. Preliminary estimates of the longest acccptable campaign length while still meets ...
Date: January 1, 2002
Creator: Li, T. K. (Tien K.); Burr, Tom; Menlove, Howard O.; Thomas, K. E. (Kenneth E.); Fukushima, M. & Hori, M.
Partner: UNT Libraries Government Documents Department

Towards better tamper&intrusion detection

Description: This presentation discusses in generic terms some of the work of the Vulnerability Assessment Team at Los Alamos National Laboratory in the area of tamper and intrusion detection. Novel security approaches are discussed. We also present preliminary results for a crude prototype of a high security ('Town Crier') monitoring system for securing moving cargo or stationary assets.
Date: January 1, 2002
Creator: Johnston, R. G. (Roger G.); Garcia, A. R. E. (Anthony R. E.) & Pacheco, A. N. (Adam N.)
Partner: UNT Libraries Government Documents Department

A Study on the Tritium Behavior in the Rice Plant after a Short-Term Exposure of HTO

Description: In many Asian countries including Korea, rice is a very important food crop. Its grain is consumed by humans and its straw is used to feed animals. In Korea, there are four CANDU type reactors that release relatively large amounts of tritium into the environment. Since 1997, KAERI (Korea Atomic Energy Research Institute) has carried out the experimental studies to obtain domestic data on various parameters concerning the direct contamination of plant. In this study, the behavior of tritium in the rice plant is predicted and compared with the measurement performed at KAERI. Using the conceptual model of the soil-plant-atmosphere tritiated water transport system which was suggested by Charles E. Murphy, tritium concentrations in the soil and in leaves to time were derived. If the effect of tritium concentration in the soil is considered, the tritium concentration in leaves is described as a double exponential model. On the other hand if the tritium concentration in the soil is disregarded, the tritium concentration in leaves is described by a single exponential term as other models (e.g. Belot's or STAR-H3 model). Also concentration of organically bound tritium in the seed is predicted and compared with measurements. The results can be used to predict the tritium concentration in the rice plant at a field around the site and the ingestion dose following the release of tritium to the environment.
Date: February 26, 2002
Creator: Yook, D.-S.; Lee, K. J. & Choi, Y-H.
Partner: UNT Libraries Government Documents Department

IMPACT OF WATER PH ON ZEBRA MUSSEL MORTALITY

Description: The experiments conducted this past quarter have suggested that the bacterium Pseudomonas fluorescens strain CL0145A is effective at killing zebra mussels throughout the entire range of pH values tested (7.2 to 8.6). Highest mortality was achieved at pH values characteristic of preferred zebra mussel waterbodies, i.e., hard waters with a range of 7.8 to 8.6. In all water types tested, however, ranging from very soft to very hard, considerable mussel kill was achieved (83 to 99% mean mortality), suggesting that regardless of the pH or hardness of the treated water, significant mussel kill can be achieved upon treatment with P. fluorescens strain CL0145A. These results further support the concept that this bacterium has significant potential for use as a zebra mussel control agent in power plant pipes receiving waters with a wide range of physical and chemical characteristics.
Date: October 15, 2002
Creator: Molloy, Daniel P.
Partner: UNT Libraries Government Documents Department

A users guide for the REBUS-PC code, version 1.4.

Description: The Reduced Enrichment Research and Test Reactor (RERTR) Program uses the REBUS-PC computer code to provide reactor physics and core design information such as neutron flux distributions in space, energy, and time, and to track isotopic changes in fuel and neutron absorbers with burnup. REBUS-PC has evolved away from the original REBUS code, which was created starting in the 1960's to study large liquid metal cooled fast breeder reactors. REBUS and REBUS-PC both model the external cycle, and are very general codes with 1D, 2D, and 3D neutronics capabilities, and with complete fuel shuffling capabilities. REBUS-PC has evolved to its present status over the past decade. While it incorporates the same neutronics capabilities from DIF3D 9.0 as does REBUS 9.0 created by the RAE Division of ANL, REBUS-PC has numerous changes and enhancements directed toward the needs of the thermal reactor analyst using WINDOWS or linux-based PC's.
Date: January 30, 2002
Creator: Olson, A. P.
Partner: UNT Libraries Government Documents Department

Examination of spent PWR fuel rods after 15 years in dry storage.

Description: Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based ...
Date: February 11, 2002
Creator: Einziger, R. E.; Tsai, H. C.; Billone, M. C. & Hilton, B. A.
Partner: UNT Libraries Government Documents Department

GDH pipe break transient analysis of the RBMK - 1500.

Description: Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. The cases of GDH impact on an adjacent GDH and its attached piping are investigated in this paper. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would not significantly damage adjacent walls or piping and would not result in a propagation of pipe failures.
Date: May 15, 2002
Creator: Kulak , R.; Marcherta, A. & Dundulis, G.
Partner: UNT Libraries Government Documents Department

Regenerative Heater Optimization for Steam Turbo-Generation Cycles of Generation IV Nuclear Power Plants with a Comparison of Two Concepts for the Westinghouse International Reactor Innovative and Secure (IRIS)

Description: The intent of this study is to discuss some of the many factors involved in the development of the design and layout of a steam turbo-generation unit as part of a modular Generation IV nuclear power plant. Of the many factors involved in the design and layout, this research will cover feed water system layout and optimization issues. The research is arranged in hopes that it can be generalized to any Generation IV system which uses a steam powered turbo-generation unit. The research is done using the ORCENT-II heat balance codes and the Salisbury methodology to be reviewed herein. The Salisbury methodology is used on an original cycle design by Famiani for the Westinghouse IRIS and the effects due to parameter variation are studied. The vital parameters of the Salisbury methodology are the incremental heater surface capital cost (S) in $/ft{sup 2}, the value of incremental power (I) in $/kW, and the overall heat transfer coefficient (U) in Btu/ft{sup 2}-degrees Fahrenheit-hr. Each is varied in order to determine the effects on the cycles overall heat rate, output, as well as, the heater surface areas. The effects of each are shown. Then the methodology is then used to compare the optimized original Famiani design consisting of seven regenerative feedwater heaters with an optimized new cycle concept, INRC8, containing four regenerative heaters. The results are shown. It can be seen that a trade between the complexity of the seven stage regenerative Famiani cycle and the simplicity of the INRC8 cycle can be made. It is desired that this methodology can be used to show the ability to evaluate modularity through the value of size a complexity of the system as well as the performance. It also shows the effectiveness of the Salisbury methodology in the optimization of regenerative cycles for such an ...
Date: August 1, 2002
Creator: Williams, W.C.
Partner: UNT Libraries Government Documents Department

Assessment of MTI Water Temperature Retrievals with Ground Truth from the Comanche Peak Steam Electric Station Cooling Lake

Description: Surface water temperatures calculated from Multispectral Thermal Imager (MTI) brightness temperatures and the robust retrieval algorithm, developed by the Los Alamos National Laboratory (LANL), are compared with ground truth measurements at the Squaw Creek reservoir at the Comanche Peak Steam Electric Station near Granbury Texas. Temperatures calculated for thirty-four images covering the period May 2000 to March 2002 are compared with water temperatures measured at 10 instrumented buoy locations supplied by the Savannah River Technology Center. The data set was used to examine the effect of image quality on temperature retrieval as well as to document any bias between the sensor chip arrays (SCA's). A portion of the data set was used to evaluate the influence of proximity to shoreline on the water temperature retrievals. This study found errors in daytime water temperature retrievals of 1.8 C for SCA 2 and 4.0 C for SCA 1. The errors in nighttime water temperature retrievals were 3.8 C for SCA 1. Water temperature retrievals for nighttime appear to be related to image quality with the largest positive bias for the highest quality images and the largest negative bias for the lowest quality images. The daytime data show no apparent relationship between water temperature retrieval error and image quality. The average temperature retrieval error near open water buoys was less than corresponding values for the near-shore buoys. After subtraction of the estimated error in the ground truth data, the water temperature retrieval error was 1.2 C for the open-water buoys compared to 1.8 C for the near-shore buoys. The open-water error is comparable to that found at Nauru.
Date: December 9, 2002
Creator: Kurzeja, R.J.
Partner: UNT Libraries Government Documents Department

Feasibility of natural circulation heat transport in the ENHS.

Description: An analysis has been carried out of natural circulation thermal hydraulics in both the primary and intermediate circuits of the Encapsulated Nuclear Heat Source (ENHS). It is established that natural circulation enhanced by gas injection into the primary coolant above the core, or the intermediate coolant above the heat exchange zone, is effective in transporting the nominal core power to the steam generators without the attainment of excessive system temperatures. Uncertainties in thermophysical properties and wall friction have a relatively small effect upon the calculated best estimate primary and intermediate coolant system temperature rises.
Date: February 14, 2002
Creator: Sienicki, J.J.
Partner: UNT Libraries Government Documents Department

Decommissioning experience from the Experimental Breeder Reactor-II.

Description: Consistent with the intent of this International Atomic Energy Agency technical meeting, decommissioning operating experience and contributions to the preparation for the Coordinated Research Project from Experimental Breeder Reactor-II activities will be discussed. This paper will review aspects of the decommissioning activities of the Experimental Breeder Reactor-II, make recommendations for future decommissioning activities and reactor system designs and discuss relevant areas of potential research and development. The Experimental Breeder Reactor-II (EBR-II) was designed as a 62.5 MWt, metal fueled, pool reactor with a conventional 19 MWe power plant. The productive life of the EBR-II began with first operations in 1964. Demonstration of the fast reactor fuel cycle, serving as an irradiation facility, demonstration of fast reactor passive safety and lastly, was well on its way to close the fast breeder fuel cycle for the second time when the Integral Fast Reactor program was prematurely ended in October 1994 with the shutdown of the EBR-II. The shutdown of the EBR-II was dictated without an associated planning phase that would have provided a smooth transition to shutdown. Argonne National Laboratory and the U.S. Department of Energy arrived at a logical plan and sequence for closure activities. The decommissioning activities as described herein fall into in three distinct phases.
Date: March 28, 2002
Creator: Henslee, S.P. & Rosenberg, K.E.
Partner: UNT Libraries Government Documents Department

EVALUATION OF BIOTIC AND TREATMENT FACTORS RELATING TO BACTERIAL CONTROL OF ZEBRA MUSSELS

Description: Testing over the last quarter has indicated the following regarding control of zebra mussels with bacterium Pseudomonas fluorescens strain CL0145A: (1) the concentration of bacteria suspended in water is directly correlated with mussel kill; (2) the ratio of bacterial mass per mussel, if too low, could limit mussel kill; a treatment must be done at a high enough ratio so that mussels do not deplete all the suspended bacteria before the end of the desired exposure period; (3) bacteria appear to lose almost all their toxicity after suspension for 24 hr in highly oxygenated water; (4) in a recirculating pipe system, the same percentage mussel kill will be achieved irrespective of whether all the bacteria are applied at once or divided up and applied intermittently in smaller quantities over a 10-hr period. Since this is the fourth quarterly report, a summation of all test results over the last twelve months is provided as a table in this report. The table includes the above-mentioned fourth-quarter results.
Date: April 30, 2002
Creator: Molloy, Daniel P.
Partner: UNT Libraries Government Documents Department

A Gamma Radiolysis Study of UO{sub 2}F{sub 2} 0.4H{sub 2}O Using Spent Nuclear Fuel Elements from the High Flux Isotope Reactor

Description: The development of a standard for the safe, long-term storage of {sup 233}U-containing materials resulted in the identification of several needed experimental studies. These studies were largely related to the potential for the generation of unacceptable pressures or the formation of deleterious products during storage of uranium oxides. The primary concern was that these conditions could occur as a result of the radiolysis of residual impurities--specifically fluorides and water-by the high radiation fields associated with {sup 233}U/{sup 232}U-containing materials. This report documents the results from a gamma radiolysis experiment in which UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O was loaded in helium. This experiment was performed using spent nuclear fuel elements from the High Flux Isotope Reactor as the gamma source and was a follow-on to experiments conducted previously. It was found that upon gamma irradiation, the UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O released 0{sub 2} with an initial G(O{sub 2}) = 0.01 molecule O{sub 2}/100 eV and that some of the uranium was reduced from U(VI) to U(IV). The high total dose achieved in the SNF elements was sufficient to reach a damage limit for the UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O. This damage limit, measured in terms of the amount of the U(IV) produced, was found to be about 9 wt%.
Date: January 24, 2002
Creator: Icenhour, A.S.
Partner: UNT Libraries Government Documents Department